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Petti IAEA - Nuclear Sciences and Applications - IAEA

Recent Accomplishments

and Future Directions in the

US Fusion Safety &

Environmental Programs

David Petti

IAEA TCM on Fusion Safety

Vienna, Austria

July 12, 2006


Outline

• STAR

• Dust

• Tritium Material Interactions and Permeation

• Fusion Safety Codes

• Risk

• Waste Management

• MFE Safety Design

• IFE Safety Design

• Summary


Dust/Debris Characterization

Home of the STAR Facility at the

Idaho National Laboratory

TFTR DIII-D C-MOD

Tore Supra

NOVA

Tritium Plasma Experiment TPE Plasma

W brush

samples

Tritium Uptake in Materials

Fusion Safety Chemical

Reactivity Experiments

STAR

A National User Facility Providing a

Valuable Resource for Researchers Across

the Nation and Around the World.

Be sample

after exposure

to ion beam

Molten Salt Tritium/

Chemistry Experiments

Tritium Infrastructure

Systems

Fusion Safety Mobilization Testing

Mo alloy samples after

exposure to air


Glovebox TCS

MS tritium exp

MS corrosion exp

Tritium SAS

STAR Floorplan Layout

Stack monitor

TPE

D-ion implantation

2LiF-BeF 2 preparation,

purification and testing

15,000 Ci tritium limit

Segregation of operations

Gloveboxes and hoods

Tritium cleanup system

Once-through room ventilation

Chemical reactivity


Star Tritium Storage and Assay System

SAS glovebox Setup

Secondary inlet

Primary outlet

U-bed-1

U-bed-2

SAS manifold with U-beds

SAS manifold and

vacuum pumps


STAR Tritium Cleanup System (TCS)

Condenser

MS water

Outlet IC

Blower

Inlet

Catalyst beds

Inlet sample loop

Control

Heat exchanger

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)


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

Timeline of TPE at LANL and INL

March 2002 - Extraction,

loading, and transport

April 8, 2002 - Depart LANL

April 10, 2002 - Arrival at INL

STAR Facility

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

Spring &

Summer 2005

Electrical

Service

re-design

& construction,

experiment

& facility

interface

assemblies

completed

Fall 2005 - Integrated systems testing initiated

December

2005 - First

plasma testing

(non-tritium)


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 on

tritium trapping

• Certification of these structures for ITER

• Use of tritium in the plasma will enable low level

measurements needed for such research


Flibe Tritium Experiment: Design and testing of

tritium handling systems and diagnostics

Flow meter

Ar D 2

T 2

Vacuum

pump

Flow meter

Flow meter

Pressure gauge

Cap

exhaust

HF trap

High temperature salt

Flibe

Conceptual layout proposed by Fukada et al.

Ni

Gas chromatograph

or QMS

or ionization chamber

dip Be if Redox control is successful

• Tritium provided in

pressurized vessel

containing D 2 /T 2 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


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 generation

ITER Key Issues: chemical reactivity, radioactivity

content, dust explosions.

Science: understand and model underlying formation

mechanisms to estimate inventories expected in fusion

ITER Dust Strategy from EDA

Demonstration

of the filtered

vacuum

collection

technique

Key scientific issues

needing resolution


TFTR

JET

Tore Supra

LHD

NOVA

Measurements of dust characteristics are an active

area of research

Particle Size Distribution, Specific Surface Area, Surface Mass Density,

Composition, Shape and Tritium Content

DIII-D

Comparison of Size Distributions

Machine

Alcator-Cmod

TEXTOR

ASDEX-Upgrade

Lower

Regions

0.66 + 2.82

0.88 + 2.63

1.58 + 2.80

27 + (-)

5-20 + (-)

2.68 + 2.89

2.21 + 2.93

8.59 + 2.67

1.12 + 1.90

CMD (μm) + GSD

Middle

Regions

0.60 + 2.35

1.60 + 2.33

1.53 + 2.80

-

-

2.98 + 2.94

3.69 + 2.81

6.31 + 2.39

0.76 + 2.03

Upper

Regions

0.89 + 2.92

1.22 + 2.03

-

3.32 + 2.94

3.59 + 3.08

8.73 + 2.09

0.90 + 1.93

R&D continues to resolve the issues

-

-

Specific Surface Area, m 2 /g

100

10

1

Specific Surface Area

0.1

fully dense C

fully dense Mo

TFTR

DIII-D

Alcator-Cmod

Tore Supra

0.01

1

ASDEX-Upgrade

NOVA

JET

ATJ graphite

10

Mo

C

100

Mean Volume-Surface Diameter, d (μm)

MVS

Median Particle Diameter (μm)

15

10

5

0

10 0

LHD

Dust Size versus Surface Mass Density

ASDEX-Upgrade

Tore Supra

CMOD

10 1

10 2

Surface Mass Density (mg/m 2 )

LHD

ASDEX-Upgrade

Tore Supra

CMOD_(98)

DIIID_(98)

10 3

DIII-D

10 4


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 in

Risk Assessment for Fusion

• Work on component failure rate data to support quantitative safety

assessment continues to be very successful.

– Initially, component failure rates were collected from handbooks

and applied to fusion.

– Now, through IEA Task 5, we collect fusion facility operating

experience data from TLK, TPL, and the former TSTA; and tokamak

data 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 assessment

of the ITER facility to identify occupational hazards

– Occupational injury rates have been collected from several US

tokamaks and large particle accelerators

– WE-FMEA method was developed to address highly hazardous

equipment failures in a fusion facility, such as high energy pipe

breaks 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 Layer

Oxidation 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-ofvacuum

accidents (LOVAs)

Beryllium oxidation rate (kg/m 2 -s)

10

10

10

10

10

0

-3

-6

-9

-12

INL92

88% dense MELCOR dust layer model

( Dust =0.7g/cm 3 , d p = 20μm)

INL fully dense

INL98 88% dense

INL Dust

5 10 15 20

10,000/T (K)


MAGARC Poloidal Field Coil Development

• The MAGARC code was recently

modified to analyze unmitigated

quench events in ITER poloidal

field (PF) magnets

• This modification include

the electrical

characteristics of the twoin-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.

Voltage (V)

0 500 1000 1500

0.00

Radial direction

0.25

Height (m)

Time 110s

Winding pair

0.50 0.75 1.00

Axial direction

Inline arcs

voltage drops

0.00 0.18 0.36 0.540.72

Width (m)

Radial direction

Axial direction


Number of arcs

MAGARC Poloidal Field Coil Application

• MAGARC code

application to

unmitigated

quench events in

ITER poloidal field

(PF) magnets

8

6

4

2

Inline

Radial

Axial

Lead voltage drop (V)

0

0 150 300

Time (s)

450 600

Quench fraction

3000

2000

1500

1000

500

0

0.8

0.6

0.4

0.2

0.0

0 150 300 450 600

Time (s)

Coil current (kA)

0 150 300 450 600

Time (s)

Melt volume (m 3 )

4.0

3.0

2.0

1.0

0.0

0.12

0.08

0.04

0 150 300 450 600

Time (s)

0.00

0 150 300

Time (s)

450 600


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 Radwaste

Management

• 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 to

commercial market if CL < 1.

• Tighter environmental controls and the political difficulty of

building new repositories worldwide may force fusion designers

to 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 fusion

components.

• Recycling of liquids and solids may generate limited amount of

radioactive waste that needs special treatment.


Blanket/Shield/Vacuum Vessel/Magnet/Structure

Volume (10 3 m 3 )

3.5

3.0

2.5

2.0

1.5

1.0

0.5

SiC

0.0

ARIES – I

1990

ARIES Project Committed to

Waste Minimization

FS

(D- 3 He)

III

1991

Tokamaks

II

1992

SiC

V V

IV

1992

RS

1996

ST

1999

SiC

AT

2000

Tokamak waste volume

halved over 10 y study period

FS

Blanket/Shield/Vacuum Vessel/Magnet/Structure

Volume (10 3 m 3 )

8

7

6

5

4

3

2

1

0

FS

UWTOR-M

24 m

1982

V

SPPS

14 m

1994

Stellarators

FS

ARIES-CS

8.25 m

2006

FS

ARIES-CS

7.5 m

2006

Stellarator waste volume

more than halved over

25 y study period


U.S. Clearance Index

Volume (10 3 m 3 )

1.0

0.9

0.8

0.6

0.5

0.4

0.3

0.1

0.0

10 12

10 10

10 8

10 6

10 4

10 2

10 0

10 -2

10 0

FW/Blkt/

BW

FW

Vacuum Vessel

10 2

80% of ARIES-CS Active Materials can be

Cleared in < 100 y after Decommissioning

Inter-Coil Structure

Steel of Bldg

Concrete of Bldg

Limit

Shld/

Mnfld

10 4

1d 1y

10 6

Time (s)

Cryostat

Not compacted, no replacements

Fully compacted with replacements

VV

Magnets &

Structure Cryostat

10 8

100y

10 10

Blanket

Shield

Vacuum

Vessel

Cryostat

2 m Bioshield

IAEA Clearance Index

Magnet

Manifolds

10 12

10 10

10 8

10 6

10 4

10 2

10 0

10 -2

10 0

FW

Vacuum Vessel

Inter-Coil Structure

Cryostat

Steel of Bldg

Concrete of Bldg

Clear

Limit

10 2

Recycle or

Dispose of

10 4

1d

10 6

Time (s)

1y

10 8

100y

10 10

Recycle or

Dispose of

B/S/VV/M

(20%)

Clear

Magnet w/o Nb Sn,

3

Cryostat & Bioshield

(80%)


Blanket

Shield

Vacuum

Vessel

Cryostat

All ARIES-CS Components can be

Recycled in 1-2 yr Using Advanced and

Conventional Equipment

Magnet

Manifolds

Recycling Dose Rate (Sv/h)

10 6

10 4

10 2

10 0

10 -2

10 -4

10 -6

10 -8

10 0

Shield

Inter-Coil Structure

Conc. of Bldg-I

Steel of Bldg-I

Time (s)

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:

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

– Strict impurity control may allow hands-on recycling.

VV

10 2

10 4

FW

SiC

1d

Cryostat

10 6

1y

10 8

10 10

Advanced

RH RH Limit Limit

Conservative

RH RH Limit Limit

Hands-on Hands-on

Limit Limit


MELCOR Code Applied to ARIES Compact

Stellarator Reactor Safety Assessment

• ARIES-CS has Dual Cooled Liquid

Lead Lithium (DCLL) Blanket

• This blanket concept employees

reduced activation ferritic steel

(RAFS) for the structure, cooled by

8 MPa helium, and a self-cooled

breeding zone cooled by flowing

PbLi.

• Helium pressurization accidents will

be a concern for this concept, with

the reactor cryostat serving as

secondary radioactivity and

pressure confinement

• Because stellarator plasmas do not

disrupt, the beyond design basis

bypass accident (BDBA) of concern

may be a heat exchanger tube

breach in the PbLi/Brayton cycle

system, leaking 10 MPa helium into

the blanket causing a blanket break

and pressurization of the vacuum

vessel

Pressure (MPa)

0.25

0.20

0.15

0.10

0.05

Vacuum Vessel

Cryostat

0.00

3600 3700 3800

Time (s)

3900 4000

Initial results for in-vessel FW helium LOCA


MELCOR Code Applied to Advance Power

Extraction (APEX) Reactor Safety Assessment

• APEX studied an advanced Ferritic

Steel FLiBe breeder Reactor

Concept

• Evaluated a confinement bypass

accident

– Total loss of site power that

induces a beyond design basis

plasma disruption the resulting

in the loss of confinement

through a heating or diagnostic

duct to an adjoining room

• Source terms included structure

activation products by oxidation,

tungsten dust from the divertor

erosion, FLiBe activation products by

evaporation, and structure tritium by

diffusion

• Conservative weather conditions and

stack release assumed

Loss-of-Vacuum Accident

VV duct isolation valves fail

and reentrant flow is

established in duct

T 2 released from

FW forms HTO

in air humidity

Aerosols form from FW oxidation

and molten salt LOCA

Reentrant flow established in

HVAC duct transports

aerosols to the environment

Aerosols transported

to and deposition in

non-nuclear room


Temperature (C)

1200

1000

800

600

MELCOR Code Applied to Advance Power

Extraction (APEX) Reactor Safety Assessment

Temperature Response

LOCA without VV cooling

LOCA with VV cooling

LOFA with VV cooling

LOFA without

VV cooling

400

0 5 10 15

Time (d)

Dose (mSv)

0.5

0.4

0.3

0.2

0.1

Site boundary dose

Flibe

0.0

0 1 2 3 4 5 6 7

Time (d)

Tritium

AFS

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.


MELCOR Code Applied to US Test Blanket

Module Safety Assessment

• Evaluate consequences to ITER

from accidents in the proposed

US DCLL Test Blanket module

(TBM)

• To date three accident

scenarios have been

investigated:

– In-vessel TBM coolant leaks

– In-TBM breeding zone

coolant leaks

– Ex-vessel TBM cooling

system leaks

• No significant impacts on ITER

safety have been identified, but

assessment is still ongoing

• All FS structures

are He-cooled by

8 MPa

• PbLi self-cooled

flows in poloidal

direction

Internal

PbLi flow

He out

He in

PbLi

concentric

inlet/outlet

pipe


S&E considerations are critical

for the success of IFE

• IFE has both radiological and toxicological hazards:

– Tritium fuel, activated structural material, activated dust, activated

coolants 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 and

radiation

• 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 but

closely linked to the future of IFE research


In the recent years there has

been great progress in IFE S&E

• In order to maximize the S&E advantages of IFE, accident

consequences must be addressed realistically

• In early studies, safety analysis tools were not very refined which

often resulted in overly conservative safety analyses and safetyimportant

design details were not available to incorporate into the

safety assessment

• We have adopted and adapted computer codes traditionally used

by MFE, and integrated them in a set of state-of-the-art

codes/libraries for IFE safety analyses

• These tools have provided the first self-consistent analysis to

understand the integrated behavior of an IFE chamber under

accident conditions

• This methodology was applied to various IFE designs and a target

fabrication facility, demonstrating that the implementation of the

Fusion Safety Standards in IFE power plant designs was achievable

HYLIFE-II

SOMBRERO


S&E activities in support of the

HAPL Program

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

• Dominant issue in accident scenario with Li chemical reactions is mobilization of

tritium and activated structural materials

Schematic of HAPL chamber using

self-cooled liquid Li blanket

Temperature (K)

MELCOR predicted temperature

evolution during Li fire

1.2E+03

1.0E+03

8.0E+02

6.0E+02

4.0E+02

2.0E+02

0.0E+00

FW

2nd wall

Back wall

Shielding

Building

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

Time (s)


1

H

3

Li

11

Na

19

K

37

Rb

55

Cs

Z-IFE RTL waste disposal rating

study: Goal is WDR < 1

4

Be

12

Mg

20

Ca

38

Sr

56

Ba

21

Sc

39

Y

57

La

22

Ti

40

Zr

72

Hf

58

Ce

23

V

41

Nb

73

Ta

59

Pr

24

Cr

42

Mo

74

W

60

Nd

WDR > 10

1 < WDR < 10

0.1 < WDR < 1

0.01 < WDR < 0.1

WDR < 0.01

Not studied

25

Mn

43

Tc

75

Re

61

Pm

26

Fe

44

Ru

76

Os

62

Sm

27

Co

45

Rh

77

Ir

63

Eu

28

Ni

46

Pd

78

Pt

64

Gd

00

Ex

29

Cu

47

Ag

79

Au

65

Tb

30

Zn

48

Cd

80

Hg

66

Dy

5

B

13

Al

31

Ga

49

In

81

Tl

67

Ho

6

C

14

Si

32

Ge

50

Sn

82

Pb

68

Er

Weekly recycling rating

Daily recycling rating

7

N

15

P

33

As

51

Sb

83

Bi

69

Tm

8

O

16

S

34

Se

52

Te

84

Po

70

Yb

9

F

17

Cl

35

Br

53

I

71

Lu

2

He

10

Ne

18

Ar

36

Kr

54

Xe


(

e

t

a

r

e

s

o

d

t

c

a

t

n

o

C

10 7

10 6

10 5

10 4

10 3

10 2

10 1

10 0

10 -1

10 4

Example of impact of safety

analyses in NIF

• Material selection is an important part of controlling worker doses

• Al-5083 was preferable to stainless steels at the “decay times” of greatest

interest for worker doses and decomissioning

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

rate by 3

5 days

10 5

10 6

Time (s)

Al-5083

SS-316

10 7

10 8

3 years

(

e

t

a

r

e

s

o

d

t

c

a

t

n

o

C

10 4

10 3

10 2

10 1

10 0

10 -1

10 -2

10 4

5 days

10 5

Concrete with no Boron

Concrete with Boron

10 6

Time (s)

10 7

3 years

10 8


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 and

tritium source terms will answer important questions for ITER and

future machines.

• Regulatory approval of ITER and the associated verification and

validation activities for our fusion safety codes and risk and

reliability methods will provide greater confidence in application of

these tools to evaluate public and worker safety of future fusion

facility designs.

• The resurgence of nuclear fission reactor construction activities

worldwide will cause increased attention to waste management

issues associated with nuclear power which in turn should help

fusion as it develops a long term waste management strategy

consistent with on-going US regulation.

• Safe and environmentally sound operation of both ITER and NIF

will be important public demonstrations of the S&E potential of

fusion.

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