Recent Accomplishmentsand Future Directions in the

US Fusion Safety &Environmental Programs

David PettiIAEA TCM on Fusion Safety

Vienna, AustriaJuly 12, 2006

Outline

• STAR

• Dust

• Tritium Material Interactions and Permeation

• Fusion Safety Codes

• Risk

• Waste Management

• MFE Safety Design

• IFE Safety Design

• Summary

Fusion Safety ChemicalReactivity Experiments

Molten Salt Tritium/Chemistry Experiments

Fusion Safety Mobilization Testing

Tritium Uptake in Materials

Home of the STAR Facility at theIdaho National Laboratory

Tritium Plasma Experiment TPE Plasma

STAR

Be sampleafter exposureto ion beam

Mo alloy samples afterexposure to air

Dust/Debris Characterization

TFTR DIII-D C-MOD

Tore SupraNOVA

A National User Facility Providing aValuable Resource for Researchers Across

the Nation and Around the World.

W brushsamples

Tritium InfrastructureSystems

STAR Floorplan Layout

15,000 Ci tritium limit

Segregation of operations

Gloveboxes and hoods

Tritium cleanup system

Once-through room ventilation

2LiF-BeF2 preparation,

purification and testing

D-ion implantation

TPE

Chemical reactivity

Glovebox TCS

Tritium SAS Stack monitor

MS tritium exp

MS corrosion exp

Star Tritium Storage and Assay System

U-bed-2

U-bed-1

Secondary inlet

Primary outlet

SAS manifold with U-bedsSAS manifold and

vacuum pumpsSAS glovebox Setup

STAR Tritium Cleanup System (TCS)

Inlet

Control

Outlet IC

Heat exchanger

Catalyst beds

MS water

Blower

Inlet sample loop

Condenser

Mole sieve bed

Tritium is now on-site at STAR• Useable tritium inventory now 1300 Ci

– 300 Ci in equimolar H2:D2:T2 calibration standard

– 1000 Ci T2 available for experiments

– shipments from SRS limited to 1000 Ci with

standard TYPE-A shipping container

Shipping Vessel (1 available)

Timeline of TPE at LANL and INL

May 1995 - First tritium plasma

experiments at TSTA

September 2000 - Final tritium

plasma experiments at TSTA

February 2001 - Begin D&D

efforts of TPE

December 2001 - Final pump

out of system; close all valves

January 2002 - Preparation

for shipment

April 8, 2002 - Depart LANL

March 2002 - Extraction,loading, and transport

December

2005 - First

plasma testing

(non-tritium)

Summer/

Fall 2002 -

Uncrate and

decon ancilliary

components

January 2003 -

Modify plans

for location of

experiment,

decide on PermaCon structure, initiate facility

design changes. June 2003- PermaCon installed,

TPE glovebox uncrated. 2004- Reassembly and

system interface design activities

April 10, 2002 - Arrival at INL

STAR Facility

Spring &

Summer 2005

Electrical

Service

re-design

& construction,

experiment

& facility

interface

assemblies

completed

Fall 2005 - Integrated systems testing initiated

Planned Research Agenda for TPE

• Study uptake, retention and permeation in PFCs

– Measurement of bulk tritium transport properties(diffusivity, solubility, dissociation/recombinatino rates)

– Monomaterials (Be, W, C)

– Mixed materials

– Bonded and/or duplex structures

– Effects of neutron dose and irradiation temperature ontritium trapping

• Certification of these structures for ITER

• Use of tritium in the plasma will enable low levelmeasurements needed for such research

Flibe Tritium Experiment: Design and testing of

tritium handling systems and diagnostics

• Tritium provided in

pressurized vessel

containing D2/T2 mixture

• Glovebox setup to contain

potential leaks

• Localized tritium cleanup will

be connected

• GC column for H isotope

separation has been tested

with tritium; works well but

needs calibration

• Develop DF/TF generator if

schedule permits

Ar D2

Flow meter

Flow meter

Flow meter

Gas chromatograph

or QMS

or ionization chamber

HF trap

exhaust

High temperature salt

Flibe

Cap

Ni

T2

Vacuum

pump

Pressure gauge

dip Be if Redox control is successful

Conceptual layout proposed by Fukada et al.

Permeation Coating Barrier Experments

Thermal Cycle Performance of He Pipe Permeation Barriers

• simulates thermal stress

degradation of permeation barrier

coatings for He pipes

• configuration matched to TBM

design for coated components

• utilize tritium for barrier

technology qualification

• external thermal cycles followed

by testing in permeation rig for

integrated effects

• in-situ thermal cycling in

permeation rig for barrier dynamic

response

Dust generationITER Key Issues: chemical reactivity, radioactivity

content, dust explosions.Science: understand and model underlying formation

mechanisms to estimate inventories expected in fusion

Demonstration

of the filtered

vacuum

collection

technique

ITER Dust Strategy from EDA

Key scientific issuesneeding resolution

0.01

0.1

1

10

100

1 10 100

fully dense Cfully dense MoTFTRDIII-DAlcator-CmodTore SupraASDEX-UpgradeNOVAJETATJ graphite

Spe

cific

Sur

face

Are

a, m

2/g

Mean Volume-Surface Diameter, dMVS

(μm)

M o

C

Specific Surface Area

Dust Size versus Surface Mass Density

0

5

10

15

100 101 102 103 104

LHDASDEX-UpgradeTore SupraCMOD_(98)DIIID_(98)

Me

dia

n P

art

icle

Dia

me

ter

(μm

)

Surface Mass Density (mg/m2)

LHD

ASDEX-Upgrade

Tore Supra

CMODDIII-D

Comparison of Size Distributions

3.32 + 2.942.98 + 2.942.68 + 2.89Tore Supra

0.90 + 1.930.76 + 2.031.12 + 1.90NOVA

8.73 + 2.096.31 + 2.398.59 + 2.67LHD

3.59 + 3.083.69 + 2.812.21 + 2.93ASDEX-Upgrade

--5-20 + (-)TEXTOR

--27 + (-)JET

1.22 + 2.031.53 + 2.801.58 + 2.80Alcator-Cmod

-1.60 + 2.330.88 + 2.63TFTR

0.89 + 2.920.60 + 2.350.66 + 2.82DIII-D

UpperRegions

MiddleRegions

LowerRegionsMachine

CMD (μm) + GSD

Measurements of dust characteristics are an activearea of research

Particle Size Distribution, Specific Surface Area, Surface Mass Density,Composition, Shape and Tritium Content

R&D continues to resolve the issues

Research Agenda for Dust

• Continued characterization in existing tokamaks

• Mobilization testing

• Chemical reactivity of dust in grooves

• Monitoring and removal evaluation

• Improved strategy for ITER

Major Accomplishments inRisk Assessment for Fusion

• Work on component failure rate data to support quantitative safetyassessment continues to be very successful.

– Initially, component failure rates were collected from handbooksand applied to fusion.

– Now, through IEA Task 5, we collect fusion facility operatingexperience data from TLK, TPL, and the former TSTA; and tokamakdata from DIII-D and JET.

– Independent data sets validate the failure rate values.

• Occupational safety is a new area for risk assessment.

– ITER IT has plans to perform a room-by-room overall assessmentof the ITER facility to identify occupational hazards

– Occupational injury rates have been collected from several UStokamaks and large particle accelerators

– WE-FMEA method was developed to address highly hazardousequipment failures in a fusion facility, such as high energy pipebreaks that have caused worker fatalities in the power industry

The Hazard Zone of the Worker Exposure -Failure Modes and Effects (WE-FMEA)

INL FSP Support of the ITER Project• The FSP is supporting the ITER Project through two Implementing Task Agreements

(ITA), established in 2004.

• U. S. ITER Fusion Safety Code Support ITA

– Provide International Team (IT) with the latest fusion versions of the MELCOR and

ATHENA codes, documentation, validation, and support and assistance at

operation of the codes.

– Delivered MELCOR 1.8.5, upgraded ice layer model for cryogenic surfaces, and

developed a beryllium dust layer oxidation model for MELCOR

– Assist the ITER IT in producing the QA documentation for MELCOR and safety

analyses for ITER’s Report on Preliminary Safety (RPrS)

• U.S. ITER Magnet Safety Task Agreement

– Update the MAGARC code to current ITER design for TF and PF coils, include

new R&D results on insulation failure behavior at elevated temperatures, apply the

MAGARC code to various ITER magnet safety studies

– Upgraded insulation and magnet parameters, implemented arc limit model,

applied MAGARC to ITER-FEAT TF and PF magnet unmitigated quench events

– Develop magnet Busbar arc capability for MAGARC

MELCOR Code Heat Structure Dust LayerOxidation Model

• A beryllium dust layer oxidation

model was developed for ITER

to simulate oxidation of a dust

layer of dust that has settled

onto a heated surface inside of

a fusion device

• This model is based on

measured oxidation reaction

rates for fully dense beryllium,

binary gaseous diffusion of

oxygen or steam into the dust

layer, and BET measured

specific surface area for

beryllium dust

• Application is for slow vacuum

vessel pressurization events

from in-vessel loss-of-cooling

accidents (LOCAs) and loss-of-

vacuum accidents (LOVAs)

1210

10

10

10

10

-

-9

-6

-3

0

Be

rylli

um

oxid

atio

n r

ate

(kg

/m2-s

)

5 10 15 2010,000/T (K)

INL Dust

INL fully dense

INL98 88% dense

INL92

88% dense MELCOR dust layer model( Dust=0.7g/cm3, dp= 20μm)

MAGARC Poloidal Field Coil Development

• This modification include

the electrical

characteristics of the two-

in-hand winding pair of the

ITER PF coils and limits

on the number of arcs that

can form in the magnet

during unmitigated quench

events based on an

energy minimization

principle.

Ax

ial directio

n

Radial direction

• The MAGARC code was recently

modified to analyze unmitigated

quench events in ITER poloidal

field (PF) magnets

voltage drops

Time 110s

0.18

0.00

0.360.540.720.25

0.00

0.500.75

1.00

500

01000

1500

Voltage (

V)

Width (m)

Height (m)

Inline arcs

Axial directionRadial direction

Winding pair

MAGARC Poloidal Field Coil Application

• MAGARC code

application to

unmitigated

quench events in

ITER poloidal field

(PF) magnets

0.0

0.2

0.4

0.6

0.8

Que

nch

frac

tion

0 150 300 450 600Time (s)

0

500

1000

1500

2000

Lead

vol

tage

dro

p (V

) 3000

0 150 300 450 600Time (s)

0 150 300 450 600Time (s)

0.0

1.0

2.0

3.0

4.0

Coi

l cur

rent

(kA

)0 150 300 450 600

Time (s)

0

2

4

6

8

Num

ber

of a

rcs Inline

RadialAxial

0 150 300 450 600

Time (s)

0.00

0.04

0.08

0.12

Mel

t vol

ume

(m3 )

Future Safety Code Activities

• MAGARC capabilities will be expanded to treat arcs in magnet

busbars by including the magnetic effects of the arc that forms

between the leads of a busbar. As part of an international

collaboration, this new capability will be validated against data that

has recently been obtained from the MOVARC experiment at FzK in

Germany

• We will be working with the ITER IT to provide the necessary quality

assurance documentation required for ITER licensing for the

MELCOR code

• We will continue in support of the licensing process for the US DCLL

TBM to complete Dossier on Safety for this TBM concept

• We will complete the safety assessment of the ARIES Compact

Stellarator and continue in support of the design activity as ARIES

takes on a new design vision

Recent Trends in RadwasteManagement

• Options:

– Disposal in repositories: LLW (WDR < 1) or HLW (WDR > 1)

– Recycling – reuse within nuclear facilities (dose < 3000 Sv/h)

– Clearance – release slightly-radioactive materials tocommercial market if CL < 1.

• Tighter environmental controls and the political difficulty ofbuilding new repositories worldwide may force fusion designersto promote recycling and clearance, avoiding geological disposal

No radwaste burden on future generation.

• There’s growing international effort in support of this new trend.

• Recycling may not be economically feasible for all fusioncomponents.

• Recycling of liquids and solids may generate limited amount ofradioactive waste that needs special treatment.

ARIES Project Committed toWaste Minimization

Tokamak waste volumehalved over 10 y study period Stellarator waste volume

more than halved over25 y study period

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

Bla

nket/

Sh

ield

/Vacu

um

Vessel/M

ag

net/

Str

uctu

re

Vo

lum

e (

10

3 m

3)

ARIES – I1990

III1991

II1992

IV1992

RS1996

ST1999

AT2000

SiCSiC

SiC

V V

FS

FS(D-

3He)

0

1

2

3

4

5

6

7

8

Bla

nket/

Sh

ield

/Vacu

um

Vessel/M

ag

net/

Str

uctu

re

Vo

lum

e (

10

3 m

3)

UWTOR-M24 m1982

SPPS14 m1994

ARIES-CS 8.25 m

2006

V FS

FS

ARIES-CS 7.5 m2006

FSTokamaks Stellarators

80% of ARIES-CS Active Materials can beCleared in < 100 y after Decommissioning

10-2

100

102

104

106

108

1010

1012

100

102

104

106

108

1010

U.S

. C

leara

nce I

nd

ex

Time (s)

1d 1y

Inter-Coil Structure

Limit 100y

FW

Vacuum Vessel

Cryostat

Steel of Bldg

Concrete of Bldg

10-2

100

102

104

106

108

1010

1012

100

102

104

106

108

1010

IA

EA

Cle

ara

nce I

nd

ex

Time (s)

1d 1y

Inter-Coil Structure

Limit

100y

FW

Vacuum Vessel

Cryostat

Steel of Bldg

Concrete of Bldg

0.0

0.1

0.3

0.4

0.5

0.6

0.8

0.9

1.0

Not compacted, no replacementsFully compacted with replacements

Vo

lum

e (

10

3 m

3)

FW/Blkt/BW

Shld/Mnfld

VV Magnets &Structure

Cryostat

Recycle orDispose ofB/S/VV/M

(20%)

Clear Magnet w/o Nb

3Sn,

Cryostat & Bioshield(80%)

2 m Bioshield

Cryostat

Blanket

Manifolds

Shield

VacuumVessel

Magnet

Recycle orDispose of

Clear

All ARIES-CS Components can beRecycled in 1-2 yr Using Advanced and

Conventional Equipment

10-8

10-6

10-4

10-2

100

102

104

106

100

102

104

106

108

1010

Recyclin

g D

ose R

ate

(S

v/h

)

Time (s)

Advanced RH Limit

Conservative RH Limit

Hands-onLimit1y1d

FW

SiCShield

Inter-Coil Structure

Steel of Bldg-IConc. of Bldg-I

VV

Cryostat

Advanced RH Limit

Conservative RH Limit

Hands-onLimit

FS-based components:

– 54Mn (from Fe) is main contributor to dose.

– Store components for few years before recycling.

– After several life-cycles, advanced RH equipments may be needed to handle shield,manifolds, and VV.

SiC-based components:

– 58,60Co, 54Mn, and 65Zn contributors originate from impurities.

– Strict impurity control may allow hands-on recycling.

Cryostat

Blanket

Manifolds

Shield

VacuumVessel

Magnet

MELCOR Code Applied to ARIES CompactStellarator Reactor Safety Assessment

• ARIES-CS has Dual Cooled LiquidLead Lithium (DCLL) Blanket

• This blanket concept employeesreduced activation ferritic steel(RAFS) for the structure, cooled by8 MPa helium, and a self-cooledbreeding zone cooled by flowingPbLi.

• Helium pressurization accidents willbe a concern for this concept, withthe reactor cryostat serving assecondary radioactivity andpressure confinement

• Because stellarator plasmas do notdisrupt, the beyond design basisbypass accident (BDBA) of concernmay be a heat exchanger tubebreach in the PbLi/Brayton cyclesystem, leaking 10 MPa helium intothe blanket causing a blanket breakand pressurization of the vacuumvessel

3600 3700 3800 3900 4000Time (s)

0.00

0.05

0.10

0.15

0.20

0.25

Pre

ssu

re (

MP

a)

Vacuum VesselCryostat

Initial results for in-vessel FW helium LOCA

MELCOR Code Applied to Advance PowerExtraction (APEX) Reactor Safety Assessment

• APEX studied an advanced FerriticSteel FLiBe breeder ReactorConcept

• Evaluated a confinement bypassaccident

– Total loss of site power thatinduces a beyond design basisplasma disruption the resultingin the loss of confinementthrough a heating or diagnosticduct to an adjoining room

• Source terms included structureactivation products by oxidation,tungsten dust from the divertorerosion, FLiBe activation products byevaporation, and structure tritium bydiffusion

• Conservative weather conditions andstack release assumed

Aerosols transported

to and deposition in

non-nuclear room

VV duct isolation valves fail

and reentrant flow is

established in duct

Reentrant flow established in

HVAC duct transports

aerosols to the environment

T2 released from

FW forms HTO

in air humidity

Aerosols form from FW oxidation

and molten salt LOCA

Loss-of-Vacuum Accident

MELCOR Code Applied to Advance PowerExtraction (APEX) Reactor Safety Assessment

Temperature Response Site boundary dose

0 5 10 15

Time (d)

400

600

800

1000

1200

Tem

pera

ture

(C

)

LOCA without VV cooling

LOCA with VV cooling

LOFA without VV cooling

LOFA with VV cooling

Major contributors AFS dose are Mn-54, Ca-45, and Ti-45,plant

isolation must occur within four weeks to stay below the 10 mSv limit.

0 1 2 3 4 5 6 7

Time (d)

Dos

e (m

Sv)

0.0

0.1

0.2

0.3

0.4

0.5

Flibe

AFS

Tritium

MELCOR Code Applied to US Test BlanketModule Safety Assessment

• All FS structures

are He-cooled by

8 MPa

• PbLi self-cooled

flows in poloidal

direction

He out

He in

Internal

PbLi flow

PbLi

concentric

inlet/outlet

pipe

• Evaluate consequences to ITERfrom accidents in the proposedUS DCLL Test Blanket module(TBM)

• To date three accidentscenarios have beeninvestigated:

– In-vessel TBM coolant leaks

– In-TBM breeding zonecoolant leaks

– Ex-vessel TBM coolingsystem leaks

• No significant impacts on ITERsafety have been identified, butassessment is still ongoing

S&E considerations are criticalfor the success of IFE

• IFE has both radiological and toxicological hazards:

– Tritium fuel, activated structural material, activated dust, activatedcoolants or coolant impurities, and activated gases

– Chemically toxic materials (i.e.: Hg, Pb)

• Energy sources that can mobilize these hazardous materials include:

– chemical energy, decay heat, pressure energy, electrical energy andradiation

• In the US, current IFE S&E activities, are focused on a few programs:

– The High Average Laser Program (HAPL)

– The Z-IFE Program

– The National Ignition Facility (NIF): not really an IFE program butclosely linked to the future of IFE research

In the recent years there hasbeen great progress in IFE S&E

• In order to maximize the S&E advantages of IFE, accidentconsequences must be addressed realistically

• In early studies, safety analysis tools were not very refined whichoften resulted in overly conservative safety analyses and safety-important design details were not available to incorporate into thesafety assessment

• We have adopted and adapted computer codes traditionally usedby MFE, and integrated them in a set of state-of-the-artcodes/libraries for IFE safety analyses

• These tools have provided the first self-consistent analysis tounderstand the integrated behavior of an IFE chamber underaccident conditions

• This methodology was applied to various IFE designs and a targetfabrication facility, demonstrating that the implementation of theFusion Safety Standards in IFE power plant designs was achievable

HYLIFE-II

SOMBRERO

S&E activities in support of theHAPL Program

• We have completed preliminary S&E assessment for the HAPL design

• Dominant issue in accident scenario with Li chemical reactions is mobilization oftritium and activated structural materials

Schematic of HAPL chamber using

self-cooled liquid Li blanket

0.0E+00

2.0E+02

4.0E+02

6.0E+02

8.0E+02

1.0E+03

1.2E+03

0.0E+00 2.0E+05 4.0E+05 6.0E+05 8.0E+05 1.0E+06

Time (s)

Te

mp

era

ture

(K

)

FW

2nd wall

Back wall

Shielding

Building

MELCOR predicted temperature

evolution during Li fire

Z-IFE RTL waste disposal ratingstudy: Goal is WDR < 1

1

H

2

He

3

Li

4

Be

5

B

7

N

6

C

8

O

9

F

10

Ne

11

Na

12

Mg

13

Al

14

Si

15

P

16

S

17

Cl

18

Ar

19

K

37

Rb

55

Cs

20

Ca

31

Ga

32

Ge

33

As

34

Se

35

Br

36

Kr

25

Mn

26

Fe

27

Co

28

Ni

29

Cu

30

Zn

21

Sc

22

Ti

23

V

24

Cr

38

Sr

49

In

50

Sn

51

Sb

52

Te

53

I

54

Xe

43

Tc

44

Ru

45

Rh

46

Pd

47

Ag

48

Cd

39

Y

40

Zr

41

Nb

42

Mo

56

Ba

81

Tl

82

Pb

83

Bi

84

Po

75

Re

76

Os

77

Ir

78

Pt

79

Au

80

Hg

57

La

72

Hf

73

Ta

74

W

67

Ho

68

Er

69

Tm

70

Yb

61

Pm

62

Sm

63

Eu

64

Gd

65

Tb

66

Dy

58

Ce

59

Pr

60

Nd

71

Lu

WDR > 10

1 < WDR < 10

0.1 < WDR < 1

0.01 < WDR < 0.1

WDR < 0.01

Not studied

00

Ex

Weekly recycling rating

Daily recycling rating

10-2

10-1

100

101

102

103

104

104

105

106

107

108

Concrete with no BoronConcrete with Boron

Contact dose rate (

Time (s)

10-1

100

101

102

103

104

105

106

107

104

105

106

107

108

Al-5083SS-316

Contact dose rate (

Time (s)

Example of impact of safetyanalyses in NIF

• Material selection is an important part of controlling worker doses

• Al-5083 was preferable to stainless steels at the “decay times” of greatestinterest for worker doses and decomissioning

• Addition of boron (0.14% by weight) to gunite shielding reduced t=5 day doserate by 3

5 days

3 years5 days

3 years

Summary• US S&E research continues to help improve fusion facility design in

terms of accident safety, worker safety, and waste disposal.

• The R&D underway and currently planned in the areas of dust andtritium source terms will answer important questions for ITER andfuture machines.

• Regulatory approval of ITER and the associated verification andvalidation activities for our fusion safety codes and risk andreliability methods will provide greater confidence in application ofthese tools to evaluate public and worker safety of future fusionfacility designs.

• The resurgence of nuclear fission reactor construction activitiesworldwide will cause increased attention to waste managementissues associated with nuclear power which in turn should helpfusion as it develops a long term waste management strategyconsistent with on-going US regulation.

• Safe and environmentally sound operation of both ITER and NIFwill be important public demonstrations of the S&E potential offusion.