ML20035F979
| ML20035F979 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 04/16/1993 |
| From: | Mike Franovich Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9304230173 | |
| Download: ML20035F979 (86) | |
Text
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UNITED STATES g
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NUCLEAR REGULATORY COMMISSION V.cic -
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,E WASHINGTON. D. C. 20555 April 16, 1993 Docket No.52-002 APPLICANT: ABB-Combustion Engineering, Inc. (ABB-CE)
PROJECT:
CE System 80+
SUBJECT:
PUBLIC MEETING OF FEBRUARY 17 AND 18, 1993, TO DISCUSS SEVERE ACCIDENT ANALYSIS FOR THE CE SYSTEM 80+ STANDARD PLANT DESIGN l
l On February 17 and 18, 1993, a public meeting was held at the Duke Engi-neering & Services (DE&S) facilities in Charlotte, North Carolina, between representatives of the U.S. Nuclear Regulatory Commission (NRC) and ABB-CE.
l The purpose of the meeting was to discuss ABB-CE's progress on resolving issues discussed in the CE System 80+ severe accidents meeting held December 2 and 3, 1993, in Windsor, Connecticut. provides a list of attend-ees. is the material presented by ABB-CE.
The first issue discussed was the structural capability of the reactor cavity under static and dynamic loads from an ex-vessel steam explosion (EVSE) event.
ABB-CE presented the results of the structural evaluation.
For static loads, the reactor cavity's pressure capacity is estimated to be at least 225 psi for the severe accident condition (design basis load combinations specified in Standard Review Plan Section 3.8.3 are not yet completed).
For the dynamic pressure capacity analysis, the cavity is modelled by including steam genera-tor supports from the cavity floor (elevation 62 ft.) to the top of the first major floor slab (elevation 90 ft. 6 3/8 in.) assuming fixed boundary condi-tions at the base, top, and outside corners of the steam generator supports with cracked concrete properties below the corbels except on the steam generator supports. A natural period (T) of 0.0114 second (87.825 Hz) was obtained.
Using a dynamic increase factor specified in Appendix C to ACI 349-85, the dynamic pressure capacity (R,) is estimated as 259 psi.
As an example, a pressure impulse of 288 psi was obtained with the ductility ratio (p) of 3, an assumed impulse duration (t ) of 0.005 seconds, and a e
square forcing function shape. Table 1 (Enclosure 2) shows the impulse l
pressures with different forcing shapes and ductility ratios.
For the CE System 80+ containment shell, ABB-CE used the minimum required yield strength of 60 ksi as the base for. SA-537 steel, Class 2 material.
Based on the 122 sample tests of SA-537 steel, the actual yield had a mean value of 69.1 ksi with a standard deviation of 3.3 ksi. Thus, the actual yield is expected to be about 15 percent higher than the minimum yield predicted.
For the purpose of establishing the containment fragility curve, ABB-CE used 2-D and 3-D models.
From the 2-D model, 5 percent probability of failure (P,) for pressure at ultimate capacity (P, the collapse load defined in Section III, Appendix II, Article 1430 of the 1SME Code) with minimum 1
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, April 16, 1993 l
required yield strength, 50 percent probability of failure to P, with 10 per-cent higher than minimum required yield strength, and 100 percent probability were assigned., with 20 percent higher than minimum required yield strength of failure to P From the 3-D model, 3 percent probability of failure was assigned to the pressure at yield (P ) of 142.42 psi using the minimum required yield required strength.
Fromthelestdata,theprobabilityoffailureat60ksiis 0.3 percent. Table 2 (Enclosure 2) shows the various pressures with proba-bilities of failure as a function of temperatures.
The pressures with 50-percent probability failure appeared high when various factors in an actual "as-manufactured" spherical containment shell, such as residual stresses, non-uniform thickness, varying values for o or geometric p
imperfections (radii of curvature, dents, peaking at welds) are taken into account.
ABB-CE stated that stresses are not allowed to go beyond the yield point l
(i.e., pressures should be controlled under 142.42 psi). The 142.42 psi was assigned as the ASME Service Level C pressure limit.
Also, the stress-strain curve with the minimum required yield strength of 60 ksi will be used for strain requirements such as at containment penetrations.
l ABB-CE also discussed the hydrogen mitigation system (HMS) and the placement l
of the 42 hydrogen igniters throughout the containment (Enclosure 3) in accordance with 10 CFR 50.34(f) requirements. ABB-CE stated that engineering l
judgment was used~ as the basis for igniter placement similar to previous work performed for the Maguire and Catawba fluclear Power Stations.
The following commitments were made during the meeting:
(1)
ABB-CE will document the design-evaluation for the location of the reactor cavity sump with respect to the EVSE scenario. ABB-CE will present options and justification for the options evaluated such as use of a low-profile pump and motor to compensate for in-core instrumenta-tion (ICI) tubes clearance, the Electric Power Research Institute Utility Requirements Document sump requirements, feasibility for use of an alternate type pump versus a wet-pit pump, use of a concrete drainage l
cover, etc.
l (2)
ABB-CE will provide a summary of the~ assumptions, calculations, and results of the reactor cavity structural analysis for radial ablation cases.
(3)
ABB-CE will respond to the axial ablation case by specifying minimum concrete / wall thickness, erosion rates, and type of concrete (basaltic, limestone). This response will be folded into the response to request for additional information (RAI) #26 from the liRC/ABB-CE severe acci-dents meeting held on December 2 and 3, 1992.
(4)
ABB-CE will update / revise the CE System 80+ structural design guidelines document to specify a lower corbel rebar pattern to support the reactor cavity structural analysis for coping with an EVSE event.
I m
. April 16, 1993 i
(5)
ABB-CE committed to provide the complete fragility curve with material uncertainty and modeling uncertainty and the material characteristics for penetration seals to ensure minimal containment leakage at higher pressures and temperature.
(6)
ABB-CE will provide written documentation explaining the basis for the i
development of the containment fragility curve.
l (a) Specifically, a discussion on the use of the 3-D model and 2-D axisymmetric model and the basis for assigning a 50-percent proba-i bility to the 110 percent of the yield stress (Sy) for the ultimate capacity calculation.
(b)
Based on discussions during the meeting, the assignment of a 50-percent probability of containment failure at 1.l*Sy appeared overly conservative and may warrant a more accurate /best estimate value, especially when considering tests of SA-537 steel have demonstrated a 15 percent greater than predicted yield point in the material. However, ABB-CE should account for as-built imperfec-tions in their revised best-estimate, as discussed previously.
(c) ABB-CE will provide a discussion on use of medium data values for i
the ultimate capacity calculations as opposed to use of minimum values.
(7)
ABB-CE will specify in CESSAR-DC Chapter 3 the selection of a high temperature material for containment penetration seals that is consis-tent with the assumptions of the probabilistic risk assessment (PRA)
(i.e., 450 *F).
(8)
In response to the RAls from the severe accident meeting held on December 2 and 3,1992, ABB-CE committed to evaluate the capability to vent the containment under overpressure conditions.
Based on this meeting, ABB-CE will expand its original evaluation and document the 4
options evaluated (e.g., purge system, hydrogen recombiner lines to annulus, backup containment spray, etc.).
ABB-CE will also cross-reference the PRA assumptions used for these overpressure sequences (e.g., use of a dedicated high-pressure /high-capacity mobile pump to supply the containment spray header, as discussed in the January 4 and 5, 1993, PRA meeting).
(9)
ABB-CE will provide a discussion on the role of the hydrogen igniter system in accident mitigation and the approach adopted in the placement of the 42 hydrogen igniters. This discussion should address' regulatory requirements of 10 CFR Part 50, CESSAR-DC Chapter 6 commitments, and the role the igniters play in the System 80+ PRA.
(10) ABB-CE will identify the minimum set of equipment / instrumentation necessary to function in a post-hydrogen burn environment.
ABB-CE will also quantify and discuss the method to quantify equipment survivability in this environment.
- April 16, 1993 (11) ABB-CE will define hydrogen deflagration environments (globally and locally), duration, and burn profiles to support equipment survivability quantification.
(12) The staff will determine how severe accident design features and insights will be treated for design certification purposes.
Specific-1 ally, the staff will determine information that should be captured in-O c
(a) Tier 1 material (design description / inspections, tests, analyses, i
and acceptance criteria)
(b) Tier 2 material (standard safety analysis report)
(13) ABB-CE will provide volumes, dimensions, and flow areas for four selected areas in the reactor cavity area.
(14) ABB-CE will provide a discussion of battery loads for severe accident equipment / instrumentation (minimum set and other necessary loads).
j Finally, ABB-CE presented an overview of the next revision andGorthcoming update to the " System 80+ Severe Accident Phenomenology and Containment Performance Report."
(Original signed by)
Michael X. Franovich, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIBUTION w/ enclosures:
Docket File PDST R/F MFranovich DCrutchfield PDR RPerch, 8H7 PShea SJLee, 7H15 MSnodderly, 801 JKudrick, 8D1 DISTRIBUTION w/o enclosures:
TMurley/FMiraglia WTravers RBorchardt ACRS (11)
WRussell TEssig JMoore, 15B18 GGrant, EDO TWambach EJorda MNBB3701 MMalloy GBagchi, 7H15 RBarrett, SH7
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LA:PDjl:AgR SC. t DSSA PM:P(JST: ADAR S
sT:ADAR PSh(5 JKudr% (
MXEfanovich:tz TEssig 04/ p &
NAME:
04/' '/93 04/i,/93 04/,L,/93 DATE:
0FFICIAL RECORD COPY: MSdM0217.MXF
l ABB-Combustion Engineering, Inc.
Docket No.52-002 cc: Mr. C. B. Brinkman, Acting Director Mr. Walter Pasedag Nuclear Systems Licensing U.S. Department of Energy ABB-Combustion Engineering, Inc.
Washington, D.C.
20585 1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations ABB-Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing i
ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 l
Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
l Washington, D.C.
20503 Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Joseph R. Egan, Esquire l
Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.
Washington, D.C.
20037-1128 Mr. Regis A. Matzie, Vice President Nuclear Systems Development ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 L
MEETING ATTENDEES FEBRUARY 17 and 18, 1993 NAME ORGANI7ATION M. Franovich NRR/PDST S. J. Lee NRR/ECGB M. Snodderly NRC/DSSA J. Kudrick NRC S. Ritterbusch ABB-CE L. Gerdes ABB-CE R. Schneider ABB-CE l
W. Pasedag DOE l
T. Oswald DE&S J. Johnson DE&S l
R. Wagstaff DE&S L. Davis DE&S T. Crom DE&S l
l l
l l
r.
,.n,,
\\
Table 1 System 80+ Reactor Cavity Dynamic Pressure Capacity Rm = 259 psi Square Forcing Function Shape To =0.003 seconds T =0.005 seconds T, / T =.003/.0114 = 0.26 T,/T =.005/.0114 = 0.44 4=1 F =259/1.4= 185 psi F =259/2.0 = 130 psi 1
i 4=3 F =259/0.65= 398 psi F =259/0.9 = 288 psi 3
3 p=5 F =259/0.5= 518 psi F =259/0.75 = 345 psi 3
3 4=10 F =259/0.35 = 740 psi F =259/0.56 = 463 psi 3
3 Triangular Forcing Function Shape T, =0.003 seconds T, =0.005 seconds T, / T =.003/.0114 = 0.26 T,/T =.005/.0114 = 0.44 p=1 F =259/0.7= 370 psi F =259/1.1= 235 psi 3
3 4=3 F =259/0.32= 809 psi F =259/0.5= 518 psi 3
3 4=5 F =259/0.23= 1,126 psi F =259/0.4= 648 psi 3
3 p=10 F =259/0.17= 1,524 psi F =259/0.28= 925 psi 3
3
+.
I Table 2 Pressures with Probabilities of Failure (P,s)
Temp.
P) o.
Pressures with P, b
d
("F)
(psi)
(psi) 3%*
5%
50%'
100%
150 167.7 57,500 142.42 156 173 189 290 153.0 52.480 129.75 142 157 172 350 149.0 51,100 126.27 138 153 167 450 142.3 48.800 120.46 132 145 159 1:
With 1.0 o of 60 ksi y
a:
From 3-D model with 1.0 o of 60 ksi for yield y
b:
From 2-D model with 1.0 o of 60 ksi for ultimate (ASME collapse load) y c:
From 2-D model with 1.1 o of 60 ksi for ultimate (ASME collapse load) y d:
From 2-D model with 1.2 o of 60 ksi for ultimate (ASME collapse load) y Comparisons of Probabilities of Failure Between Advanced Reactors Level C 1.0a, 1.la.
1.15a.
- 1. 2a.,
B, m
s s
ABWR 2.2%
27.8%
50.0%
61.3%
70.1%
0.16 Sys 80+*
3.0%
5.0%
50.0%
100.0%
?
AP600' 100%
?
a:
From the test, P, at 1.00 0.3%
=
7
Table 2 Pressures with Probabilities of Failure (P,s)
Temp.
P) a.
Pressures with P, b
d
(*F)
(psi)
(psi) 3%*
5%
50%*
100%
i 150 167.7 57,500 142.42 156 173 189 290 153.0 52,480 129.75 142 157 172 350 149.0 51,100 126.E7 138 153 167 450 142.3 48,800 120.46 132 145 159 1:
With 1.0 o of 60 ksi y
a:
From 3-D model with 1.0 o of 60 ksi for yield y
b:
From 2-D model with 1.0 o of 60 ksi for ultimate (ASME collapse load) y c:
From 2-D model with 1.1 o of 60 ksi for ultimate (ASME collapse load) y d:
From 2-D model with 1.2 o of 60 ksi for ultimate (ASME collapse load) y Comparisons of Probabilities of Failure Between Advanced Reactors 4
Level C 1.0a.
1.10 1.15o.
1.20 8,
ABWR 2.2%
27.8%
50.0%
61.3%
70.1%
0.16
)
Sys 80+'
3.0%
5.0%
50.0%
100.0%
?
AP600*
100%
?
a:
From the test, P, at 1.0a = 0.3%
y
i l
l i
System 80+ Containment Performance Containment Ultimate Pressure Values I
Penetration Effects Stress-Strain Curve i
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Q,' ' ' ' ' ' ' ' ' ' ' ; l ll ' l l l,' l,'//,,',7 SYSTEM 80+
THREE DIMENSIONAL MODEL OF STEEL CONTAINMENT i
- System 80+ J w
Service Level C Containment Pressure Values With Temperature Degradation 3 - D Model TEMPERATURE E
STRESS STRAIN PRESSURE
(*F)
(psi)
(psi)
X Y
Z (psig) 150 29.07E+06 Sy = 57,500 0.0015 0.0009
-0.0010 142.42 290 28.35E+06 Sy = 52,480 0.0014 0.0008
-0.0010 129.75 350 28.00E+06 Sy = 51,100 0.0014 0.0008
-0.0010 126.27 450 27.50E+06 Sy = 48,800 0.0014 0.0008
-0.0009 120.46
Ultimate Containment Pressure Values With Temperature Degradation Axisymmetric Model TEMPERATURE E
STRESS STRAIN PRESSURE
( F)
(psi)
(psi)
X Y
Z (psig) 150 29.07E+06 Sy = 57,500 0.0012
-0.0011 0.0014 156 1.1 x Sy = 63,250 0.0013
-0.0012 0.0016 173 1.2 x Sy = 69,000 0.0014
-0.0014 0.0017 189 290 28.35E+06 Sy = 52,480 0.0011
-0.0010 0.0014 142 1.1 x Sy = 57,728 0.0012
-0.0012 0.0015 157 1.2 x Sy = 62,976 0.0013
-0.0013 0.0016 172 350 28.00E+06 Sy = 51,100 0.0011
-0.0010 0.0013 138 1.1 x Sy = 56,210 0.0012
-0.0011 0.0015 153 1.2 x Sy = 61,320 0.0013
-0.0012 0.0016 167 450 27.50E+06 Sy = 48,800 0.0010
-0.0010 0.0013 132 1.1 x Sy = 53,680 0.0011
-0.0011 0.0014 145 1.2 x Sy = 58,560 0.0013
-0.0012 0.0016 159 i
4
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o 60 50 Stress (ksi) 40 -
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a 70 --
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l System 80+ Reactor Cavity Performance l
Concrete Ablation 1
Sump Design and Location j
I Refueling Cavity Seal i
cavity Static and Dynamic Pressure Capacities 1
l f
I i
1 1
l i
System 80+ Reactor Cavity Dynamic Pressure Capacity Em = 259 psi Square Forcing Function Shape T, =0. 003 seconds To =0.005 seconds T,/T =.003/.0114 = 0.26 Ta / T =.005/ 0114 = 0.44 p=1 F =259/1.4= 185 psi F =259/2.0 = 130 psi 3
3 i
p=3 F =255/0.65= 398 psi F =259/0.9 = 288 psi 3
3 p=5 F =259/0.5= 518 psi F =259/0.75 = 345 psi 3
3 l F =259/0.56 = 463 psi p=10 F =259/0.35 = 740 psi 3
3 l
Triangular Forcing Function Shape Ta =0. 003 seconds T, =0.005 seconds Ta /T =.003/.0114 = 0.26 To / T =.005/.0114 = 0.44 l
p=1 F =259/0.7= 370 psi F =259/1.1= 235 psi 3
3 p=3 F =259/0.32= 809 psi F =259/0. 5= 518 psi i
3 3
p=5 F =259/0.23= 1,126 psi F =259/0.4= 648 psi 3
3 p=10 F =259 /0.17= 1,524 psi F =259/0.28= 925 psi 3
3 I
1
)
System 80+ Reactor Cavity Modal Analysis ANSYS Finite Element Computer Code Analysis Model Cavity modelled including steam generator supports from cavity floor, elevation 62'-0",
to top of first major floor slab, elevation 90'-6 3/8".
Fixed conditions at base, top and outside corners of steam generator supports.
Used cracked concrete properties below corbels except on steam generator supports.
Modal Analysis Results First breathing mode-87.825 hert (0.0114 seconds)
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g.,
STATUS OF SYSTEM 80 + SEVERE ACCIDENT ISSUES FEBRUARY 18,1993
__m_
6 OVERVIEW 0
CONTAINMENT CHALLENGES
--EARLY
-- LATE O
ANCILLARY ISSUES
--COMPOSITION / MASS OF EJECTED MELT
--RV FAILURE SIZE
--PRE-VB RCS PIPING FAILURE 4
l
4 COMPOSITION / MASS OF EJECTED CORIUM FROM RV i
I 0
LOWER PLENUM EJECTED MASS IS OF SIGNIFICANCE FOR
EX-VESSEL STEAM PRODUCTION l
EX-VESSEL STEAM EXPLOSION 0
AMOUNT OF MASS AND COMPOSITION OF MASS IN LOWER PLENUM PRIOR TO VB BASED ON SEVERE ACCIDENT SCALING METHODOLOGY FOR DCH; o
APPROACH IS TYPICALLY CONSERVATIVE IN THAT IT OVERESTIMATES PURE ZIRCALOY PHASE l
l o
SASM RESULTS ARE BASED ON A SYNERGISTIC ASSESSMENT OF DEGRADED CORE PROCESSES PREDICTED BY A RANGE OF COMPUTER CODES AND INCLUDES VARIOUS LOWER HEAD FAILURE CONDITIONS i
o DISTRIBUTION MADE SYSTEM 80+ SPECIFIC BY ADJUSTING RESULTS TO A PLANT WITH HIGHER ZR CONTENT AND LOW AG-IN-CD.
l ALL BUNDLE STRUCTURAL STEEL CREDITED IN SASM WAS l
CONVERTED TO PURE ZR FOR SYSTEM 80+
--- ALL AG-IN-CD WAS TAKEN TO BE STEEL ALL MASSES WERE RATI0ED UP BY TO REFLECT THE DIFFERENCE i
i IN THE TOTAL CORE MASS e
o, J
IABLE 4.1.1-1.
M(LT COMPOSITION AT VB FOLLOWING A STATIOW BL ACLOUT
+
BASED ON R[f(Rf WCE 4.61 l
HIGN PRESSURE V11H VB HlCN PRESSURE WITH LOW PRESSURE TAltuRE CREEP TAltURf PENElRATION TAftuRE TOTAt MASS EJFCIED (%)
40.0 34.0 63.0 i
FRACTION Of EJECTED 0.815 1.00 0.445 MASS IN A MOLIEN STATE l
MOLTEh MASS COMPOSITION
(*CS)
STEEt 16,000.
8,500.
28,700.
ZR 15,000.
15,000.
13,000.
00, 24,000.
34,300.
5.000.
i
- 2ro, 0.00 0.00 0.00 SOLIO MASS COMPOSITION i
frcS)
STEEL 0.00 0.00 0.018 j
l U0, 12,200.
0.00 59,100.
i
- Zro, 0.000 0.00 0.000
- TOTAL CORE MASS = 168,000 KGs (including tower support structure) 4 j
4-5 4
., - -. ~.. - - -, _... - - - -,
- 1 4
RV LOWER HEAD FAILURE SIZE O
THE RV LOWER HEAD FAILURE SIZE INFLUENCES
EVSE ROCKET FAILURE 0
CALCULATIONAL METHODOLOGY BASED ON ICI TUBE EJECTION FOLLOWED BY CORIUM ABLATION OF THE HOLE.
--- CALCULATIONS ARE CONSISTENT WITH SASM AND MAAP METHODOLOGY
--- MAXIMUM HOLE SIZE PRIOR AT POINT OF COMPLETE CORIUM EJECTION l
EQUAL TO 0.5 FT2 0
FAILURE MECHANISM 0F TUBE EJECTION STRONGLY SUPPORTED BY INEL FAILURE MAPS j
WHEN RCS IS BELOW RELIEF VALVE SETPOINT i
i 0
AT RELIEF VALVE PRESSURES THIS FAILURE FAILURE MODE COMPETES WITH A i
GLOBAL FAILURE MODE.
HIGH PRESSURE FAILURES PRIOR TO VB ARE LOW PROBABILITY <
1% OF PDS TIME DEPENDENT FAILURE IN CREEP IS UNKNOWN AND THE HOLE i
SIZE NECESSARY TO DEPLETE THE RV 0F CORIUM MAY BE SMALL SINCE THE CORIUM DISCHARGE TIME IS SHORT' i
Appendix F; Figure F.2 RV Failure Area vs. Time RV Failure Area (Ft2) 0.6 i
i
.4....4
.i.
l i
0.4
- + -
i-
- - + - -
1
.+
..t.+
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..9...
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0.2 1-i 0.1
+-
+-
+--
V
[
O
\\
l 0
1 2
3 4
5 6
7 8
9 10 Time (seconds)
RVFAIL1; HTC=140000, PRV 2500 d
)
l
5 Appendix F; Figure F.3 RV Failure Area vs. Time RV Failure Area (Ft2) 0.5
.i-0.4
-~
/
j'/
0.3 r-I O.2 1-
~ -
0.1
-+
~
i 0
)
0 1
2 3
4 5
6 7
8 9
10 I
Time (seconds) l RVFAIL2; HTC 84000, PRV 1100 i
i i
2
b a-a
_A&.-._
--_..LA a.
w.
J
_._.2-Aa-.
e-h
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DEPRESSURIZATION OF RCS PRIOR TO VB MECHANISMS 1.
OPERATOR OPERATION OF SDS 2.
INDUCED RCS PIPING FAILURE 3.
RCP SEAL FAILURE i
4 4
4 5
. -.m.-.
DEPRESSURIZATION PRIOR TO VB OPERATOR OPERATION OF SAFETY DEPRESSURIZATION SYSTEM MOST HIGH PRESSURE SEQUENCES RESULT FROM EITHER INACTION OF THE OPERATOR TO ACTUATE THE SDS IN TIME TO AVERT CORE DAMAGE OR A SMALL LOCA WHERE OPERATION OF SDS DOES NOT CHANGE CORE DAMAGE SEQUENCE.
THEREFORE, THE SDS IS AVAIALBLE TO THE OPERATOR DURING THE ACCIDENT PROGRESSION AND IS EXPECTED TO BE USED TO DEPRESSURIZE THE RCS PRIOR TO V8 5
i i
a l
DEPRESSURIZATION PRIOR TO VB THERMALLY INDUCED RCS PIPING FAILURE O
PRIMARY GUIDANCE BASED ON EXPERT ELICITATION FOR NUREG-1150 0
THERMALLY INDUCED RCS PIPING FAILURE CAN RESULKT FROM 1.
CREEP FAILURES IN THE SURGE LINE DUE TO FORCED FLOWS OCCURRING DURING SDS/PSV OPERATION 2.
CREEP FAILURES IN THE HOT LEG / HOT LEG N0ZZLE OF FIELD WELDS ASSOCIATED WITH NATURAL CIRCULATION AND COMPLICATED BY l
STRATIFIED HOT STEAM FLOWS.
3.
STEAM GENERATOR TUBE FAILURES i
0 EXPERT ELICITATION CONCLUDED THAT PROBABILITY OF INDUCED RCS PIPING FAILURE AT HIGH PRESSURE (PSV SETPOINT) WAS.95 4
1 i
a
DEPRESSURIZATION PRIOR TO VB APPLICABILITY OF EXPERT JUDGEMENT TO SYSTEM 80+
0 FOR CONSIDERATION OF SURGE LINE INDUCED FAILURES 1.
SURGE LINES ARE OF SIMILAR MATERIALS / THICKNESS 2.
NORMALIZED SRV FLOW IS LARGER THAN THAT FOR SURRY 3.
SYSTEM 80+ HAS HIGHER ZR CONTENT: GREATER CHEMICAL ENERGY INPUT (HIGHER STEAM / HYDROGEN TEMPERATURES)
SYSTEM 80+ IS AT LEAST AS LIKELY TO UNDERGO SURGE LINE FAILURE AS WOULD SURRY 0
FOR HOT LEG /N0ZZLE FAILURE 1.
SYSTEM 80+ HAS HIGHER ZR CONTENT:
GREATER DRIVING FORCE FOR NAT. CONVECTION AND HIGHER TEMPERATURES 2.
SYSTEN 80+ HOT LEG WALL IS THICKER AND CONSTRUCTED OF CS.
SURRY HOT LEG IS THINNER (2.5 VS 4.4 INCHES) AND CONSTRUCTED OF STAINLESS STEEL.
THE SURRY HOT LEG IS THINNER AND CONSTRUCTED OF CS.
3.
IF THE FAILURE OCCURS IN THE HOT LEG N0ZZLE THE LARGER DRIVING FORCE WILL OFFSET THE THICKER MATERIAL.
l
DEPRESSURIZATION PRIOR TO VB 1
APPLICABILITY OF EXPERT JUDGEMENT TO SYSTEM 80+
4.
IF THE FAILURE OCCURS IN THE HOT LEG, THE THICKER MATERIAL IN THE SYSTEM 80+ HOT LEG WILL HAVE SIMILAR CREEP BEHAVIOR AS THE THINNER STAINLESS STEEL DUE TO DIFFERENCES IN THE CREEP CURVE AND MATERIAL PROPERTIES.
5.
FIELD WELDS ARE EXPECTED TO BE AS LIKELY TO FAIL FOR EACH DESIGN SYSTEM 80+ FEATURES WHILE DIFFERENT FROM THE REFERENCE PLANTS HAVE SIMILAR FAILURE PROBABILITY t
l 4
i l
Time to rupture CHours) g 0~
10-2 10-'
10 10 10 2
O - i i 111111l 1 i l11111l l l1lillij i l 1 l lllll l l l ll164 0 i
a o
I l
o-O O
~
3 i
O O
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0
[>
- en o
-3 U
_ o
- 3 0 to g
O
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em O
2 m %
-4 C
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+
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,+
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=e
- W m
@~l I liIllii l I Ilillll I l I l lllll l l l l lllll l l l lllti
D 6
4 EARLY CONTAINMENT CHALLENGES 0
HYDR 0 GEN COMBUSTION 0
DIRECT CONTAINMENT HEATING 0
IN-VESSEL STEAM EXPLOSIONS 0
EX-VESSEL STEAM EXPLOSION 0
ROCKET INDUCED CONTAINMENT FAILURE 4
4 4
L h
HYDROGEN COMBUSTION 0
DETERMINISTIVE CHALLENGE BASED ON COMBUSTION OF 100 % OF THE ACTIVE CLADDING (SECY-90-016) 0 METHODOLOGY 1.
ADIABATIC ISOCHORIC COMPLETE COMBUSTION (AICC) INITIATED FROM THE HIGHEST POINT ON THE FLAMMABILITY CURVE WITH THE CONSISTENT HIGHEST PRESURE REALISTIC CONSIDERATION OF MATERIAL SPECIFIC HEATS PEAK BURN PRESSURE = 102 PSIA
( MIXTURE IS BELOW 8 V/0)
ESTIMATE IS CONSERVATIVE SINCE COMPUSTION WILL NOT BE COMPLETE BELOW 8 V/0 AND AICC PREDICTIONS TYPICALLY ARE HIGH BY 10% UNDER COMPLETE COMBUSTION CONDITIONS 0
AT 290 F LEVEL C STRESS ALLOWS A 145 PSIA INTERNAL PRESSURE O
CONCLUSION HYDROGEN BURNS DO NOT POSE A SERIOUS THREAT TO CONTAINMENT INTEGRITY i
I
Figure 4.1.3-2 System 80+ Combustion Potential 14, i
115 l
I I
incomplet 110 Combustion i
O Possible i
x Combinations I
105 y i
Probable z
S
$ 10 Pressure 1
i 3g g
O Incipient 100g I
Flammable CO I
8 Mixture b
8
["
i tu I
Z 2
1 5
CC 3
90 O
I i
6~-
No Cohbustion 85 I
i N
i 4
80 0
5 10 15 20 25 30 35 40 45 50 55 60 VOLUME % STEAM; x 10-2 l
- 100 % Active Clad + Post-Burn Pressure M Flammability Limit I
i
______ _...._.__._ _ _. _.___.-. ~.___ _ ~,-...
DIRECT CONTAINMENT HEATING l
0 DCH REPRESENTS A SYNERGISTIC THREAT TO CONTAINMENT
- 1. BLOWDOWN
- 2. DIRECT HEATING
- 3. HYDROGEN COMBUSTION
- 4. METALLIC OXIDATION 0
IET TESTS INDICATE THAT IS STRUCTURES CAN RESTICT MIXING GLOBAL MIXING THE MAIN CONTAINMENT DCH IS A MODIFIED HYDR 0 GEN THREAT 0
METHODOLOGIES USED
- 1. ARSAP BOUNDING APPROACH l
l
- 2. NRC APPROACH (PILCH METHOD) t i
9 i
i l
i
.. ~ -..
DIRECT CONTAINMENT HEATING ARSAP BOUNDING APPROACH 0
METHDDOLOGY ASSUMES:
1.
SINGLE CELL MODEL 2.
ALL DEBRIS EJECTED INTO THE CONTAINMENT IS FULLY OXIDIZED AND GIVES UP ALL ITS INTERNAL AND CHEMICAL REACTION ENERGY TO THE CONTAINMENT.
3.
MODEL ASSUMES 100 DISPERSAL OF EJECTED DEBRIS TO UPPER COMPARTMENT (SYSTEM 80+ DISPERSAL EXPECTED TO BE VERY MUCH LESS THAN 10%)
4.
ALL AVAILABLE HYDROGEN IS BURNED.
VERY CONSERVATIVE APPROACH:
FOR A BEST ESTIMATE MELT EJECTION (BASED ON NUREG -1150 MEDIAN) PEAK PRESSURE < 110 PSIA FOR A 40% MELT EJECTION (SASM) PEAK PRESSURE BETWEEN 130 AND 145 PSIA FOR A 60% EJECTION WITH 50% OF DEBRIS INTO UPPER COMPARTMENT = 125 PSIA
Figure 4.1.1-2 Bounding Estimates of System 80+ Containment Responses to DCH 250 Q
..+
d
$150 I--
a e
~~
G I
o 3
5!
e1,oo
[8%
/
50 --
v-is
,s Typical Corium dispersal M d.ian C o riu m dispersal
~
i NUREG-1150 y
Y l
I i
I O
1.0 0.8 0.6 0.4 0.2 0.0 Fraction of Core Ejected O1 Core release f rom RCS not credible
-=- D ry Cavity No RCS Water
@ System 80+ Ultimate Pressure Capability
-t-D ry Cavity RCS Water
@ System 80+ Stress Level C
-% Full Cavity RCS Water as Best estimate Dry cavity with RCS water e Upper bound DCH loading based on Ref. 4.3 ossumptions (D ry cavity no RCS water 50% dispersal of elected m elt
e e
e DIRECT CONTAINMENT HEATING CONTINUED OTHER COMBINATIONS FOR DCH ARE CONSIDERED IN THE PROBABILISTIC EVALUATION.
~
PROBABILISTIC EVALUATION SHOWS A CONDIDTIONAL PROBABILITY OF CF 0F 0.014 GIVEN A HIGH PRESSURE IN THE RCS AT TIME OF VB l
1 i
I 1
..m
IN VESSEL STEAM EXPLOSION 0
IVSE IS EVALUATED PROBABILISTICALLY ONLY 0
FOLLOWS GUIDANCE OF SERG 0
THIS IN-VESSEL ISSUE IS CONSIDERED GENERIC AND THEREFORE IS APPLICABLE TO SYSTEM 80+
0 NO SIGNIFICANT THREAT TO CONTAINMENT FAILURE l
l I
)
I 4
4 m
i RAPID STEAM GENERATION 0
PRESSURIZATION FOLLOWING VB WITH MELT EJECTION INTO A DEEP WATER POOL 0
PRESSURIZATION BASED ON STEAM PRODUCTION CASED BY THE TRANSFER OF ALL CORIUM STORED ENERGY INTO THE WATER POOL ALONG WITH A 50% ZR-WATER l
REACTION 0
BOUNDING CALCULATION ASSUMING:
- 1. MAXIMUM INITIAL CONDITIONS DUE TO INSTANTANE0US DISCHARGE OF INITIAL RCS INVENTORY INTO CONTAINMENT (DESIGN MAXIMUM = 67 PSIA)
(ACTUAL CALCULATED PRESSURE INCLUDING HEAT SINKS SHOW THIS PRESSURE AT THE TIME 4
OF VB TO BE UNDER 50 PSIA.
2 65% OF TOTAL CORE MASS AT AN INITIAL TEMPERATURE OF 2800 K i
- 3. STEAM BLOWDOWN OF RCS
- 4. NO CREDIT FOR PASSIVE HEAT SINKS DURING THE RAPID STEAM GENERATION PROCESS 0
MAXIMUM PRESSURE POSSIBLE IS 98 PSIA FOR SELECTED SEQUENCES.
O CONCLUSION:
NO SERIOUS DETERMINISTIC THREAT TO CONTAINMENT PROBABILISTIC THREAT LESS THAN.1% OF PRESSURIZATION CHALLENGES l
i
. _ _. _ _ _ -=
EX-VESFEL STEAM EXPLOSIONS (EVSES)
A i
0 POTENTIAL CONCERN: A SHOCK WAVE INDUCED CAVITY COLLAPSE FAILS INDIRECTLY FAILS CONTAINMNET VIA GROSS MOTIONS OF RCS t
0 METHODOLOGY i
1.
DETERMINISTICALLY ESTABLISH FRAGILITY CURVE AND LOADING TO SHOW THE LEVEL OF CONTAINMENT THREAT THAT CAN BE TOLERATED BY THE CAVITY.
2.
DEMONSTRATE THAT IN THE LOW LIKELIHOOD THAT CAVITY COLLAPSE OCCURS THE STRUCTURE IS SUFFICIENTLY ROBUST TO PREVENT AN UNIS0LABLE CONTAINMENT FAILURE.
4
)
2 2
t LEVEL OF EVSE CONTAINMENT THREAT 0
AP600 / HENRY APPROACH MAXIMUM STEAM EXPLOSION LIMITED TO A PEAK LOCAL PRESSURE OF 1450 PSIA.
THIS PRESSURE DECREASES INVERSELY AS THE SQUARE OF THE DISTANCE FROM THE 2
INTERACTION ZONE FOR A TYPICAL PULSE DURATION, THE MAXIMUM IMPULSE FOR SYSTEM 80+ IS APPROXIMATELY 1.25 PSI-SEC 0
METHOD OF MOODY i
USED IN GE SUBMITTAL.
HEAT GENERATION LIMITED BY ENHANCED FILM BOILING.
EVSE LOADING VERY LOW.
0 EQUIVALENT TNT METHODOLOGY 1.
CONVERT THERMAL ENERGY ASSOCIATED WITH MASS OF CORIUM AND SUPERHEAT TO AND EQUIVALENT CHARGE OF TNT (KINETIC ENERGY CONVERSION EFFICIENCY OF
.015 TO.03).
2.
USE HETHODOLOGY FOR ESTIMATING TNT DEPTH CHARGE LOADINGS (FROM COLE UNDERWATER EXPLOSIONS) 3.
IMPULSES CALCULATED FOR CORIUM MASSES FROM 500 TO 60,000 LBM i
i
ANL STEAM EXPLOSION WORK 4
ANL STUDIES SHOW THAT CAVITY LOADING WILL DEPEND UPON THE LEVEL OF PRE-MIXING IF SIGNIFICANT PRE-MIXING OCCURS (AS IS ANTICIPATED VIA CODES SUCH AS TEXAS) EVSE LOADS WILL BE MODEST.
TECHNOLOGY IS ASSUMPTION DRIVEN.
CONSERVATIVE ASSUMPTIONS CAN YIELD UNREALISTIC RESULTS.
ANL TASK IS TO ESTABLISH
- 1. WORK FUNCTION FOR A NOMINAL STEAM EXPLOSION (THERM 0DYMAIC EQUI. TOOL) i
- 2. PROPAGATE THE RESULTS INTO A TWO DIMENSIONAL COMPRESSIBLE FLUID DYNAMICS TOOL (ALICE)
WHEN PRE-MIXING IS HIGH LOADINGS WILL BE LOW.
l WORK IS STILL IN PROGRESS.
i i
i t
i
CONSEQUENCES OF CAVITY COLLAPSE CAVITY COLLAPSE DUE TO EVSE IS UNLIKELY.
SHOULD THIS OCCUR THE CONSEQUENCES ON CONTAINMENT INTEGRITY WILL BE SMALL.
RV SUPPORT CAN BE ESTABLISHED EVEN IF IS ALL LOWER CAVITY WALLS (MAKING UP THE CAVITY BASEMAT STRUCTURE) ARE REMOVED.
STOPS / KEYS PREVENT ROTATION AT THE STEAM GENERATOR.
THUS, CONSEQUENCE TO STEAM LINES SMALL.
RCS CONNECTING PIPING IS SUFFICIENTLY FLEXIBLE S0 THAT IT EITHER WILL NOT FAILURE IF RCS CONNECTING PIPING FAILS, IT WILL LIKELY FAIL IN-CONTAINMENT SINCE THE CONNECTING PIPING IS WEAKER THAN THE PENETRATION.
THUS, THE PROBABILITY OF A CAVITY FAILURE CASCADING INTO A CONTAINMENT FAILURE IS SMALL i
h d
PROBABILISTIC ASPECTS OF EVSE EVSES WERE PROBABILISTICALLY EVALUATED.
BASED ON i
(1) NOMINAL LOWER HEAD FAILURE,
(2) EVSE'S EFFICIENCIES OF 1.5 AND 3 PERCENT (3) CAVITY FRAGILITY CURVE THE CONDITIONAL PROBABILITY OF AN EVSE PRODUCING A CONTAINMENT FAILURE FAILURE IS ON THE ORDER OF 1 %.
4 1
s t
i
ROCKET INDUCED CONTAINMENT FAILURE ISSUE TREATED PROBABILISTICALLY CONSIDERED:
-- RV FAILURE SIZE
-- RV RESTRAINTS
-- KINETIC ENERGY REQUIRED TO REACH AND FAIL SHELL CONCLUSION GIVEN A HIGH PRESSURE CD STATE, A ROCKET INDUCED CONTAINMENT FAILURE IS LESS THAN ONE IN ONE MILLION i
I 9
4
~
LATE CONTAINMENT FAILURE 0
OVERPRESSURE FAILURE VIA STEAMING l
0 OVERPRESSURE FAILURE WITH NON-CONDENSIBLES 0
BASEMAT EROSION INDUCED CONTAINMENT FAILURE 0
OVERTEMPERATURE FAILURE O
LATE HYDROGEN BURN 4
1 b
.. -.. ~...,
OVERPRESSURE FAILURE VIA STEAMING OVERPRESSURE FAILURE IS EXPECTED FOR SCENARIOS WITHOUT SPRAYS.
USING NOMINAL MASSES AND PROPERTIES THE TIME TO REACH SERVICE LEVEL C CONDITIONS VARIES BETWEEN 50 AND 65 HOURS FOLLOWING ACCIDENT INITIATION t
a i
1
j J
d FIC QE 4 2_.111.,
T TO CONTAINMENT FAILURE FOR VARIOUS SEVERE ACCIDENT 3
(' CHALLENGES (CFS ACTUATED)3 9
~
4 TRANSIENT PDS' TIME TO TIME TO TIME *N) a CORE LEVEL C ULTIMATE UNCOVERY PRESSURE FAILURE (HR)
(HR)
PRESSURE (HR) j STATION BLACKOUT 11g 1,9 64 75 STATION BLACKOUT WITH BATTERY 242 12.1
> 65 DEPLETION TOTAL LOSS OF FW W/O FEED AND 243 1.25 62 72 BLEED J
MEDIUM I4CA WITHOUT HPSI 1
0.47 50 60 SMALL LOCA WITHOUT HPSI 201 1.05 65 K Phnt Amy swta (% rCch, rt g pea) 4-117 e
.-,.,v__., _ _..
,,.,,,.._-,.,.,...._,r,.
e.c..--..
y-_
OVERPRESSURE FAILURE WITH NONCONDENSIBLES NONCONDENSf3LE OVERPRESSURE FAILURE IS A SUBSET OF COVERPRESSURE FAILURE THIS SITUATION OCCURS WHEN THERE IS LIMITED WATER FOR STAEMING AVAILABLE AND NO SPRAYS APPROXIMAELY 2% OF THE DRY SCENARIOS CAN PRODUCE AND OVERPRESSURE FAILURE PRIOR TO A BASEMAT MELT-THROUGH
'd e
i i
l 1
{
4 1
-er,
s BASEMAT EROSION DETERMINISTIC FAILURE EVALUATION BASED ON 3 FT EROSION TO MINIMUM POSITION OF STEEL LINER AT SUMP: ISSUE TIME TO ERODE 2.5 FEET OF CONCRETE EVALUATION STUDIED BASEMAT EROSION IN THE PRESENCE OF WATER METHODOLOGIES:
ABB/CE:
MAAP ANALYSES WITH FCHF REDUCED FROM.1 DOWN TO.01 1100 KW/M2 DOWN TO 70 KW/M2 ANL:
CORCON MOD 3 GROSS EROSION AND EROSION AT SUMP i
i i
l
O RESULTS OF MAAP STUDIES i
NOMINAL RESULT IS THAT MAX EXPECTED EROSION IS.28 FT AND WILL SUBSIDE IN 6 HOURS FOR LIMESTONE /CS CONCRETE (.29 FT AND 11.5 HOURS FOR RASALTIC CONCRETE)
HEAT FLUXES DOWN TO 150 KW/M2 GIVE SUCCESS FOR L/CS CONCRETE (3 FEET IN 25 HOURS)
BELOW 150 KW/M2 CONCRETE EROSION IS AGGRESSIVE (3 FEET OF EROSION IN 17 i
HOURS) l 4
S
9 PRELIMINARY CORCON MOD 3 RESULTS MECHANISTIC CALCULATION AT 24 HOURS AFTER ATTACK LS/CS CONCRETE : 1.85 FT EROSION BASALTIC CONCRETE: 2.85 FT I
1 SUMP WILL ERODE 25 INCHES (CONSIDERABLE RADIAL EROSION)
PRELIMINARY MECHANISTIC CALCULATIONS ARE ENC 0URAGING e
t
(
4 i
i a
i
. ~,..,
.,-re-w-w-
,,., - - ~
a CONCLUSIONS 1.
THE CONTAINMENT SHELL IS PROTECTED FOR A CONSIDERABLE TIME PERIOD.
BEST ESTIMATE PREDICTIONS USING MAAP SUGGEST FULL ARREST BEFORE
.3 FT EROSION.
2.
SUCCESSFUL DEBRIS COOLING CAN BE PROVIDED FOR HEAT FLUXES DOWN TO ABOUT 200 KW/M2 FOR ALL CONCRETES 3.
FOR EROSION CONSIDERATIONS LIMESTONE / CS IS THE PREFERRED CONCRETE, FROM A PROBABILISTIC STANDPOINT BOTH ARE EQUIVALENT 1
. ~
4 OVERTEMPERATURE FAILURE i
ISSUE:
SEAL DEGRADATION AT HIGH CONT.
TEMPERATURES CAN CAUSE LOSS OF CONTAINMENT INTEGRITY ITAAC ISSUE TO SELECT APPROPRIATE MATERIAL.
THIS FAILURE MAY BE ELIMINATED.
i 1
i 1
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m
.m.A
_d AA.A e.
%&a.a A n4 m_m_&
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A
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6.
b=
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h h
LATE HYDROGEN BURN DETERMINISTIC H2 THREAT CONSIDERED IN EARLY BURN (100% ACTIVE CLAD)
PROBABILISTIC EVALUATION CONSIDERS HYDROGEN LEVELS TO 150% OF ACTIVE CLAD.
AT 150 % OXIDATION LEVELS PEAK PRESSURE EQUALS ABOUT 140 PSIA (CONDITIONAL FAILURE PROBABILITY FOR THESE SEQUENCES OF ABOUT 3%.
f i
i i
i
/'
l CONCLUSIONS 0
ALL SIGNIFICANT CONTAINMENT THREATS HAVE BEEN REVIEWED.
0 THERE ARE NO "BEST ESTIMATE" DETERMINISTIC THREATS THAT ARE OF A CONCERN TO CONTAINMENT INTEGRITY 0
PROABILISITIC EVALUATIONS SHOW THAT EXTREME EVALUATIONS OF SEVERAL THREATS WILL RESULT IN A POTENTIAL CONTAINMENT FAILURE.
QUANTIFICATION OF THESE THREATS IN LEVEL II STUDIES ARE EXPECTED TO SHOW CONDITIONAL CONTAINMENT FAILURE PROBAILITIES WITHIN THE SECY-90-016 GOALS OF 0.10.
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