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Phenomenological Advances of Severe Accident Progression
R.O. Gauntt, Reactor Analysis and Modeling Department
Sandia National Laboratories Albuquerque, New Mexico
Dept 6762 [email protected]
SOARCA Objectives
• Perform a state-of-the-art, realistic evaluation of severe accident progression, radiological releases and offsite consequences for important low likelihood accident sequences
• State of Art in– Understanding and modeling of physics and
phenomena– Fidelity and realism in system representation
Timeline of Nuclear Safety Technology Evolution
Timeline of Nuclear Safety Technology Evolution
Optimistic
Guarded
Pessimistic
Risk Informing Regulation- Modernization, NUREG-1465
License Amendments and Extension- MOX, High Burnup- Plant aging
Emergency Response PlanningAdvanced Reactors
- AP1000, ESBWR, US-EPRNGNP - HTGR, VHTGR, H2 EconomyGNEP - Fast Burner Reactor, Reprocessing
NRC
Chicago Critical Pile
Atomic Energy Act of 1954
WASH 1400
AEC
1982 SandiaSiting StudyTID 14844
source term
NUREG 1465alternative
source term
Emerging Issues.....
WASH 740
NUREG-0772NUREG-0956
MELCOR Development Advances in Core Heatup and Fission Product Release
Ag release model added – Phebus Tests – important for iodine chemistry- Important to agglomeration
B4 C oxidation model added (PWR) – QUENCH Tests, Phebus FPT-3
Fission product specie/volatility
modified (Cs2 MoO4 ) – Phebus Tests –
affects RCS deposition
Fuel failure criteria expanded via control
function – Phebus tests – affects
hydrogen generation and melt progression
BWR failure criteria expanded
Quench-reflood modeling – QUENCH tests – quench front not necessarily water level
RN Package expanded to allow analysis of FP release from mixed
MOX/LEU core
(French VERCORS and RT tests)
MELCOR Development Activities Late Phase Melt Progression
Curved lower head description added to COR (BH integration) including 2-D heat transfer
Core periphery baffle and formers added to COR package – allows TMI-like side wall melt release
Molten pool / crust model added to core and lower head regions (TMI-2)
Molten pool stratification into light and heavy layers in core
and lower head (RASPLAV,MASCA)
Natural convection heat transfer in circulating melt
pools
Fission product partitioning between melt phases
(RASPLAV,MASCA)
Insights from OECD LH program in creep-rupture
models for lower head
Phebus Program Benchmarking with MELCOR
• Integral tests• Prototypic fuel
– Fission heating– Pre-irradiated fuel
• Verification of– Fuel damage– Melt progression– Hydrogen
generation– Fission product
release and transport
– Deposition in RCS
– Containment behavior
Fuel Heatup and Degradaion
• Fission heat in FPT-1 initially drives thermal response
• Steam oxidation transient drives FP release and fuel melting
400
900
1400
1900
2400
2900
0 5000 10000 15000 20000
time [sec]
Tem
pera
ture
[K]
TCW6
TCW8
TCS 31
TC33
TU2R70TC2R70TSH170TSH270TCW6TCW8TCS31TC33
Oxidationtransient
Cladding Oxidation and Hydrogen Generation
• Steam oxidation models verified by calculated thermal response and by calculated hydrogen generation
Fission Product Release and Source Term
• ORNL Booth model captures kinetics well• CORSOR-M over-predicts early release
FPT-1 Class 2 Release - Cs
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1
5000 7500 10000 12500 15000 17500 20000
time (sec)
Rel
ease
Fra
ctio
n
ORNL BoothdataCORSOR-M
MELCOR Aerosol Mechanics - MAEROS
• MAEROS sectional model of Gelbard– 10 sections [.1 - 50 μm]– Condensed FP vapor sourced into
smallest section• Particles grow in size
– Agglomeration– Water condensation
• Particle fallout by gravitational settling• Particle deposition processes
– Thermophoresis– Diffusiophoresis– Brownian motion
• PWR Ag aerosol release from control rods modeled– Significant aerosol mass– Affects agglomeration, growth and
fallout• Cs chemisorption in RCS modeled
– Iodine from CsI revolatilizes when reheated
0.1 1 10 100aerosol diameter [μm]
sect
ion
mas
s
0.1 1 10 100aerosol diameter [μm]
sect
ion
mas
s
Aerosol size distribution evolves in time, depending on sources, agglomeration and removal processes
Time 1
Time 2
VANAM M3 Containment Aerosol Behavior Test
• Multi-compartment containment model– Steam and helium
injection– Aerosol injection
• Validation of MELCOR containment and aerosol mechanics models– Compartment
flows– Temperatures
and pressures– Aerosol transport
and deposition
Aerosol Depletion in Dome
Time [h]
18 20 22 24 26 28 30
Aero
sol C
once
ntra
tion
[kg/
m3 ]
10-7
10-6
10-5
10-4
10-3
10-2
184 Original Nodaliza184 Modified NodalizaExperiment185 Modified Nodaliza
TMI-2 Assessments
• Accident code assessments against TMI-2 ongoing in international comparison exercises
• MELCOR models for melt progression significantly refined based on TMI-2 and experiments
Realism in Modeling
• Early work required considerable abstraction in representation of plant– Computer limitations on problem size– Problem “dumbed down” to permit analysis
• Current emphasis is on realism– Spatial resolution (node size for resolving
physics)– Accurate representation of plant
• Rooms and compartments• Heat structures• Transport time and deposition
Reactor Severe Accident Modeling Evolution of the State-of-the-Art
Circa 1985 Circa 1998 Present FutureCirca 1990 Circa 1995
Timeline for Evolution of MELCOR Modeling Practices
• NUREG-1150• Basis for
NUREG-1465 revised source term
• AP-600 design certification
• ESBWR design certification
• AP-1000 design certification
• Begin SGTR • Finish SGTR • MOX source terms
• High burn-up source terms
• Emerging user needs
Example Regulatory Applications
Modeling TechniquesSTCP
CV210SG Tubes
(Up)
CV200Hot Leg
CV215SG Tubes
(Down)
CV220Pump
SuctionCV240Cold Leg
To Vessel(downcomer)
From Vessel(upper plenum)
CV450
FL440FL401
CV440
CV437
CV
435
Acc
um
FL441
CV460
C
V41
7 CV416
FL43
1
FL450
CV465
FL454
FL471
Aux Feed
FL462PORV/ADV
FL463
CV470
CV4
80
CV281(steam line/turbine)
CV001(environment)
FL48
5
Reactor Vessel
FL442
FL453
FL48
3
FL486
FL40
0
CV401
FL40
5
CV440FL442 FL443
CV420
C
V41
3 CV414
CV421
CV402
CV411 CV415
FL402
FL492
FL41
0FL
493
FL49
1
FL49
6
CV
409
CV
412
FL415
FL416FL46
5
FL49
5
FL4
33F
L494
CV425
CV426
FL41
8 FL428FL427FL417
FL452
SRV
Steam Line
MSIV
SeePressurizer
Detail
CV
450
FL440FL401
CV440
CV437
CV
435
Acc
um
FL441
CV460
C
V417
CV
416
FL4
31
FL450
CV465
FL454
FL471
Aux Feed
FL462PORV/ADV
FL463
CV470
CV
480
CV281(steam line/turbine)
CV001(environment)
FL48
5
Reactor Vessel
FL442
FL453
FL4
83
FL486
FL40
0
CV401
FL40
5
CV440FL442 FL443
CV
420
C
V413
CV
414
CV421
CV402
CV411 CV415
FL402
FL492
FL41
0FL
493
FL49
1
FL49
6
CV
409
CV
412
FL415
FL416F
L465
FL4
95
FL43
3FL
494
CV425
CV426
FL41
8 FL428FL427FL417
FL452
SRV
Steam Line
MSIV
SeePressurizer
Detail
CV
450
FL440FL401
CV440
CV437
CV
435
Acc
um
FL441
CV460
C
V417
CV
416
FL4
31
FL450
CV465
FL454
FL471
Aux Feed
FL462PORV/ADV
FL463
CV470
CV
480
CV281(steam line/turbine)
CV001(environment)
FL48
5
Reactor Vessel
FL442
FL453
FL4
83
FL486
FL40
0
CV401
FL40
5
CV440FL442 FL443
CV
420
C
V413
CV
414
CV421
CV402
CV411 CV415
FL402
FL492
FL41
0FL
493
FL49
1
FL49
6
CV
409
CV
412
FL415
FL416F
L465
FL4
95
FL43
3FL
494
CV425
CV426
FL418
FL428FL427FL417
FL452
SRV
Steam Line
MSIV
SeePressurizer
Detail
CV160Upper Head
CV140Core Region
CV110Lower Plenum
CV130Core
Bypass
CV110Downcomer
CV150Upper Plenum
CV160Upper Head
CV140Core Region
CV110Lower Plenum
CV130Core
Bypass
CV110Downcomer
CV150Upper Plenum
160
116
126
136
146 154
716
724
736 746 756
701
101
110
UpperHead
114 124
117 127 137 147 157
100
Dow
ncomer
134 144
156
114
112 122 132 142153
152
124 134 144 154199
156
172 182193
192
Bypass
718
726
738 748
LowerPlenum
758768 778 788 798
714
722
766 776 786 796
762 772 782 792752712
760 770 780 790750720
732
740
742
734
710 720 730 740 750 701
Ring 1 Ring 2 Ring 3 Ring 4 Ring 5
110
164 174 184 194
162
116 126 136146
152706
710
728
730
744 754794784774764
153
112 122 132 142
160
116
126
136
146 154
716
724
736 746 756
701
101
110
UpperHead
114 124
117 127 137 147 157
100
Dow
ncomer
134 144
156
114
112 122 132 142153
152
124 134 144 154199
156
172 182193
192
Bypass
718
726
738 748
LowerPlenum
758768 778 788 798
714
722
766 776 786 796
762 772 782 792752712
760 770 780 790750720
732
740
742
734
710 720 730 740 750 701
Ring 1 Ring 2 Ring 3 Ring 4 Ring 5
110
164 174 184 194
162
116 126 136146
152706
710
728
730
744 754794784774764
153
112 122 132 142
• MELCOR 1.8.6• Core design
details• FP Chemistry,
Impaction, other
2.0
SOARCA
core platedetail, etc.
Modeling Detail Minimizing Abstraction
Old Nodalization
CV055Dome
CV020SG Cubicles
CV041Pressurizer
Cubicle
CV050Lower Dome
CV005Basement
CV010Cavity
Enhanced Nodalization
CV41
CV's20,30,40
CV55
CV50
0.56 m
7.21 m
25.83 m
32.84 m
CV42
CV5CV5CV10
ReactorVessel
Pres
suriz
er
Ste
am G
ener
ator
PRT
16.85 m
Approx. Grade = 12.13 m
MELCOR Ref.Elevations
Grade
CV850, Auxiliary Building
CV800
LHSI
Spray Pumphouse
Auxiliary BuildingSafeguards Building
Containment
Typically model entire plant, including all potential radiological release paths (and therefore potential fission product retention mechanisms)
Modeling of Plant Buildings
160
116
126
136
146 154
716
724
736 746 756
701
101
110
UpperHead
114 124
117 127 137 147 157
100
Dow
ncomer
134 144
156
114
112 122 132 142153
152
124 134 144 154199
156
172 182193
192
Bypass
718
726
738 748
LowerPlenum
758768 778 788 798
714
722
766 776 786 796
762 772 782 792752712
760 770 780 790750720
732
740
742
734
710 720 730 740 750 701
Ring 1 Ring 2 Ring 3 Ring 4 Ring 5
110
164 174 184 194
162
116 126 136146
152706
710
728
730
744 754794784774764
153
112 122 132 142
Old Nodalization Enhanced NodalizationCV160
Upper Head
CV140Core Region
CV110Lower Plenum
CV130Core
Bypass
CV110Downcomer
CV150Upper Plenum
Model Upgrades - Vessel Nodalization -
Natural Circulation Modeling
Westinghouse PWR Core Plate
• Weight of core material mass• Engineering stress formulae used (e.g.
Roark)• Failure based on exceeding yield stress at
temperature• Sequential failure of multiple supporting
structures treated
CL
Lower coreplate
Flow mixer plate
Core plate Support Columns
1
Level
5
4
3
2
1
6
Ring 2 3 4 5
PLATEG (Level 6)
PLATEG (Level 5)
COLUMN (Level 4)
PLATE (Level 3)
Lower core plate
Flow diffuser plate
Core plate support columns
Core support forging
Impact of Modeling Advances
• Accident progression generally significantly more slowly developing
• Significantly more fission product retention in RCS and containment
• More time for mitigation of accident• Significantly reduced source terms• Even SGTR bypass source term less than
traditionally thought owing to RCS rupture in containment
Summary• Releases much delayed compared to SST-1
– Mostly many hours to 10’s compared to ~1– Due to detail modeling of reactor systems and operator
mitigations
• Releases to environment much less than SST-1– Detailed transport treatment and deposition accounting– Delayed release enhances natural depositions
• Plant models account for – Improvements in modern plant design– Important operator actions
• Peer review and uncertainty quantification work next