Phenomenological Advances of Severe Accident Progression

R.O. Gauntt, Reactor Analysis and Modeling Department

Sandia National Laboratories Albuquerque, New Mexico

Dept 6762 rogaunt@sandia.gov

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