With contributions from the FNS
Jan. 22-23, 2013
FusionEnergySystemsStudies

Fusion Nuclear Science Pathwa
Gaithersburg, MD, December 3,This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore
National Security, LLC, Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
LLNL-PRES-612712

Fusion Nuclear Science Facility (FNSF) Overview,
Physics Assump<ons and Opera<ng Space
C. E. Kessel and the FESS Team
J. Blanchard, A. Davis, L. El-Guebaly, L. Garrison, N. Ghoniem, Y. Huang,
P. Humrickhouse, Y. Katoh, A. Khodak, S. Malang,E. MarrioO, N. Morley, J. Rapp,M. Rensink, S. Reyes, T. Rognlien, A. Rowcliﬀe, S. Smolentsev, P. Snyder, M. Tillack,
P. Titus, L. Waganer, G. Wallace, P. Wilson, S. Wukitch, A. Ying, K. Young, and Y. Zhai
American Nuclear Society, Technology of Fusion Energy, August 22-25, 2016

The Fusion Energy Systems Studies Team is
Examining the Fusion Nuclear Science Facility
What does an FNSF have to accomplish?
How do we measure the FNSF progress for fusion development?
How does the FNSF accomplish its mission?
What is the pre-requisite R&D needed for an FNSF? What does
the FNSF require from our program to succeed?
How does an FNSF ﬁt in the larger fusion development program?
What cri<cal insights about this facility can be uncovered, impacts
of assump<ons, technical choices and philosophies,…?

The FNSF must ﬁll the tremendous gap between ITER
and DEMO by providing the break-in to the fusion
nuclear regime
First strongly
burning plasma

Demonstrate rouHne
Power Plant Ops
NoPowertechnicalPlantgaps

ITERFNSFDEMOMax damage

3 dpa

37-74 dpa

100-150 dpa150+ dpa

Max plasmapulse

500-3000s

1-15 days

15-365 days

365+ days~0~ 1.0

1.05+

1.05

Tblanket, Tcool,exit

285C, 150C

550C, 650C

550C, 650C

550C, 650C

Materials

316SS, CuCrZr,Be, W, H2O,
SS304, SS430

TBR

RAFM, PbLi, He, SiC-c,Borated-RAFM, W,
baini<c steel

A nearer term facility like the FNSF requires a numberof technical philosophies/approaches to be deﬁned/
explored
Facility missions and metrics – what progress does the facility make on the pathway to a powerplant
Physics strategy – how do we choose plasma parameters, what’s their impact
Long term power plant relevance – design choices are made to keep the scien<ﬁc/technology
development on track, avoid diversions that do not contribute to the power plant vision
Minimal, Moderate, and Maximal FNSF
Qualiﬁca<on requirements to install a component/material in the FNSF – fusion neutron exposure
to the dpa level, highly integrated non-nuclear tes<ng…..plasma-vacuum systems are not consistent
with “cook and look” approach to FNS, remote-maintenance
FNSF program plan – phases, material/temperature/dpa evolu<on, opera<on and maintenanceBlanket (divertor/launchers) choices and tes<ng strategy – provide the process by which we testand advance fusion core components, and backups
Hot Cell – how do we access and process the informa<on from the FNSF opera<on
Pre-FNSF R&D – how do we see the R&D evolu<on to prepare us for the FNSF, design and opera<onThe Plasma Dura<ons Required in the FNSF is a Large
Leap Compared to Present/Planned Tokamaks
Before the FNSF, must combine
ultra-long pulse linear plasma facili<estokamak conﬁnement experiments at shorter pulses
high heat ﬂux facili<esadvanced predic<ve simula<on capabilityβNPower Plant

Take advantage of the DD phase of FNSF to extend pulse
lengths

6ACT1

5KSTAR

4
3
2

Range ofpowerplants

Presentfacili<esJT-60SAEAST

DEMOFNSFITER

ACT2Pulse length, s100

101102

103

104

105

1 day

106

107

2 weeks

Physics Strategy for the FNSF Regime of longpulse, 100% non-induc<ve, burning
plasma!"#$%&'()%%&'!*+,+-)./,0"'1%#23#2'Pursue βN < no wall limit to accomplish mission, but install
appropriate feedback or other capability to exceed no wall limit
à by how much?

6,/5'!!'!'7

DIII-D expt

4
34!i2

fNI ~ 90%1

Plasma current is driven 100% non-induc<vely in ﬂaOop,
however, a solenoid provides rampup assistance and ﬂaOop
feedback
Peak heat ﬂux tolerated in the divertor < 10 MW/m2, whilepursuing high heat ﬂux design/material solu<ons and 2D SOL/
divertor plasma simula<ons
Pursuing high toroidal ﬁeld in the plasma, targe<ng LTSCadvances

2G = !NH89/q95

ELECTRONTEMPERATURE (eV)
0.3
ITER Q=5

2.0steady-state target

0.2
0.1
0.0
12.5
10.0
7.5
5.0

2.5

1101005000

PECCD (MW)
10.5

DIII–DNATIONAL FUSION FACILITY

1.0
0.8
0.6
0.4
0.2

PNBI (MW)

1.0

2.5
0.0

8' ()/' <*' ,+-(MA/m2)

Operate below the Greenwald density limit n/nGr < 1, but not
rely on low values to enhance CD

00.5
0.4

Ver<cal Posi<on (m)Install passive stabilizers and feedback coils to provide higherplasma elonga<on, signiﬁcantly expanding opera<ng space

8' 9,:' *<' #;!N23Flat-plate

40.0
0.2
0.0

Time (s)DN divertor

Greenfield, EX/1-2, Mon. p.m.

High radia<on divertor
solns
Mi<gated disrup<onsPlasma Performance and Dura.on in DIII-D and JT-60U
Looking at Experiments for Guidance
JT-60U

JT-60U

DIII-D

DIII-D

DIII-D

DIII-D

βN

2.4

1.7

3.5*

> 3.5

2.0

3.1-3.4*

τﬂaOop/τCR

2.8

2.7

2.0~1.5>2~ 0.4-1.0

q95

4.5

~86.7

5.5-6.54.7

5.0-5.5

fBS

45%

80%

40-50%

50-60%

~60%

fNI

90%

100%

75%

~ 100%

80-100%

H98

1.0

1.7

1.0

1.6

qmin

~ 1.5

1.5

~1.0

~ steadystatesteadystateà steadystate,oﬀ-axis
EAST and KSTAR will soon contributeNB

à SShybrid,
hi rot1.3> 1.2-1.3
1.4

QH-mode,no ELMs

steadystate*u<lize ac<ve error ﬁeld correc<on, plasma rota<on, βN ~ 1.15 x βNno wall
Addi<onal experiments on JT-60U, DIII-D, AUG have 1) approached and exceeded densitylimit, 2) high radiated power in the plasma and divertor, 3) avoiding or ac<vely suppressedNTMs, 4) low plasma rotaHon, and 5) PFC materials

Systems Code Iden<ﬁca<on A = 4R, m

4.80

Large scans over R, BT, q95, βN, Q, Zeﬀ,n/nGr<jTF> = 15 MA/m2
fdiv,rad = 90% (λpowFundamenski)

κX, δX

2.2, 0.63

IP, MA

7.87

BT, BTcoil, T

7.5, 15.85<jTF>, MA/m2

15 MA/m2

Filters for solu<onsβN < 2.6*
qdivpeak < 10 MW/m2
Nwpeak > 1.5 MW/m2
BTcoil < 16 T (LTSC)

βNth, βNfast

2.2, 0.23

q95

6.0

H980.99

fBS

0.52
2.43

IB Radial build from neutronics:
ΔFW/blkt = 50 cm
ΔSR = 20 cm
ΔVV = 10 cm
ΔLT shield = 23 cm
Δgaps = 20 cm

Zeﬀn/nGr

0.90

n(0)/<n>, T(0)/<T>

1.4, 2.6

Pfusion, Prad,core, Prad,div, Paux,
MW

517, 60, 160, 130

Q, Qengr

4.0, 0.86

ηCD, A-m2/W

0.2 (assumed)<Nw>, Nwpeak, MW/m2

1.18, 1.77

qdivpeak (OB, IB), MW/m2

10.7, 3.9

*examining beneﬁts of RWM
feedback to raise this toward 2.8-3.2

The FNSF Would Be Smaller Than a DEMO Plant, toReduce Cost and Facilitate a Break-in Program
These devices do not all use the samelevel of assump<ons/goals as the FNSF
Low Temp Superconducting Tokamak

Conﬁgura<on for the FNSF study:- Conven<onal aspect ra<o (= 4)- Conserva<ve tokamak physics basiswith extensions to higher
performance (βN < 2.6)

3000

- Helium cooling in blanket, shield,
divertor, and vacuum vessel- Focus on DCLL blanket concept withbackup concepts (HCLL, HCCB/PB)- Net electricity is NOT a facility
target, but electricity genera<on
can be demonstratedFusion Power, MW- 100% non-induc<ve plasma current- Low temperature superconduc<ngcoils, advanced Nb3Sn

operating space constrained byΔIBbuild = 1.23 m to TF coil
NWOB,peak > 1.5 MW/m2
βNtot < 2.6
qdivpeak < 12 MW/m2
BTcoil < 16 T

2500
K-DEMO

2000
1500

FDF-CuARC-HTSC

JA DEMO

FNSF

1000
5000

EU DEMOo

2.0

4.0

ITER
CFETR
6.0

8.0

Major radius, m

R = 4.8 m

10.0

What is the actual maximum heat ﬂux tolerable in thedivertor, what can be developed from engr design andplasma physics op<miza<on

To achieve this we know that
Engineering design
Plasma material interac<ons
Divertor/SOL plasma physics
are all required to reach a solu.on
What if we can not reach the 10 MW/
m2 that we desire…..

peak divertor heat flux, MW/m2

This represents the maximum
capability of the plate jet divertor
design….beyond this the thimble jet
design is usually required

0.85 MW/m2

Our present assump<on for the peakallowed divertor heat ﬂux is 10 MW/m2<Nw> < 1.15 MW/m2

Taking a radia<ng divertor as
the reference for FNSF
βN < 2.6BTcoil < 16 Tfdiv,rad = 0.9

qdivpeak < 10 MW/m2
7.5 MW/m2
5.0 MW/m2fdiv,rad = 0.95

ave neutron wall load at plasma, MW/m2

R = 4.8 m

Greenwald density ratio, n/nGr<Nw> < 1.15 MW/m2

auxiliary power, MW

peak divertor heat flux, MW/m2

H98(y,2) energy confinement multiplier

2D SOL analysis indicates:
Fully detached radiates ~ 100%
ITER-like divertor radiates ~ 75%

0.85 MW/m2

We assume a radiated power frac<onPdiv,rad/PSOL of 90% in systems analysis

fusion power, MW

What is the reliably
achievable radiated power
frac<on in the divertor?

βN < 2.6BTcoil < 16 T
qdivpeak < 10 MW/m2fdiv,rad = 0.95
0.90 (ref )
0.85
0.80
0.75
ave neutron wall load at plasma, MW/m2

The Program on the FNSF Deﬁnes It, Not Its Opera<ng Point
He/HDDDTDTDTDTDTPowerPlant

1.5

2-3

2.5

4.2

4.2

5.9

5.9

40 FPY

1.78

1.78

1.78

1.78

1.78

2.25

10-50

1525

35

353585Plasma
pulselength,days

Up to1012

51010310

Plasmadutycycle,%33-95

3367

91

9595

100

71926

37

37 or
74

100-15
0

RAFM
400C

RAFM ODS
450C

RAFM ODS(NS) 500C

RAFM ODS(NS) 5500C

RAFM ODS(NS) 5500C

Yrs
Neutronwall load,
MW/m2
Plasmaon-<me,% /year

10-25

Neutrondamage,dpa
blanket

RAFM
400C

RAFM
400C

Plasma pulse
extension
1 hr to 10 days

23 years of DT opera<ons, 8.4 years of neutron exposure
Higher NW, faster plasma pulse development, and eﬃcientmaintenance/plasma opera<on distribu<on can reduce years

The FNSF Program is being studied further toaddress a range of issues
1) Can the DD Phase provide enough discharge <me ranging from 1hr to 10 day pulse
lengths, u<lize higher diagnos<c coverage
2) Provide higher or lower neutron wall loads would result in short or longer years toreach a dpa level3) Low opera<ng temperatures of the early DT phases require BOTH faster coolant
ﬂow and lower fusion power, longer <me to reach dpa target4) Desire to reach longest plasma dura<ons early in the program, rather than spend
the whole program progressively extending the plasma pulse length…s<ll arriving at~7 dpa at the end of the phase5) Maintenance <mes associated with speciﬁc tasks (planned maintenance)
Ex-vessel inspec<onIn-vessel inspec<on
Minor maintenance ex-vessel
Minor maintenance in-vessel
Major in-vessel maintenance (sector removal, 16 sector removal)
à This should lead to a reduc<on in total opera<ons <meComponents in fusion
core would be evolved
and tested in the FNSF
We have concentrated on the blankets,
but there are others that may have ates<ng sequence…..materials,temperatures, design, etc.

Divertor

IB and OBBlankets

RF Launchers, TBMs,DiagnosHcs

Blanket Tes<ng
DCLL 400C RAFM (some aretaken for autopsy)
DCLL 400C RAFM/ H&CD
DCLL 450C RAFM (higher T)
DCLL 450C RAFM GII (next phase
T and FS)DCLL 400C RAFM/ MTM
DCLL 400C RAFM/ TBM-HCLL
DCLL 400C RAFM/ TBM-HCCB(PB)
DCLL 400C RAFM/ Diagnos<cPhase X – part 1

Phase X – part 2

Phase X – part 3

S-1

DCLL 400C RAFMDCLL 400C RAFM
– R1

DCLL 400C RAFM
– R1S-2

DCLL 400C RAFMDCLL 400C RAFMDCLL 400C RAFM
– R2S-3-H/CD*

DCLL 400C RAFM
– LH/EC

DCLL 400C RAFM
– LH/EC

DCLL 400C RAFM
– LH/EC

S-4-MTM*DCLL 400C RAFMDCLL 400C RAFMDCLL 400C RAFM

S-5

DCLL 400C RAFMDCLL 400C RAFMDCLL 400C RAFM

S-6

DCLL 400C RAFMDCLL 400C RAFMDCLL 400C RAFM
– R2

S-7

DCLL 400C RAFMDCLL 400C RAFMDCLL 400C RAFM

S-8

DCLL 400C RAFMDCLL 400C RAFM
– R1

DCLL 400C RAFM
– R1

S-9DCLL 450C RAFMGenIIDCLL 450C RAFMGenIIDCLL 450C RAFMGenIIS-10-H/CD

DCLL 400C RAFM– NB

DCLL 400C RAFM– NB

DCLL 400C RAFM– NB

S-11-H/CD

DCLL 400C RAFM IC

DCLL 400C RAFM IC

DCLL 400C RAFM IC

S-12

DCLL 450C RAFMDCLL 450C RAFMDCLL 450C RAFM

S-13-H/CD

DCLL 400C RAFM NBDCLL 400C RAFM NBDCLL 400C RAFM
– NB

S-14

DCLL 450C RAFMGenIIDCLL 450C RAFMGenIIDCLL 450C RAFM
GenII

S-15TBM/HCCBDCLL 400C RAFM /HCCBDCLL 400C RAFM /HCCBDCLL 400C RAFM /HCCBS-16TBM/HCLL*DCLL 400C RAFM /HCLL

DCLL 400C RAFM /HCLL

HCLL 400C RAFM /HCLL….

Divertor Tes<ng, must ﬁt
into the allocated envelope
What will be the preferred W or divertor
material?
W or W-alloy
W/X composites
Wf/Wm composites???
Will there be variants like RAFM?
Structure & armor

Flat plat fully detach
Fdiv,rad ~ 100%

Magne<c geometries
Temperature ranges
Taken from Snead, 2016

W/RAF laminate(Garrison)

FZJ

X-divertor, KDEMO
Covelle, Univ Texas

WC in Fe matrix (Álvarezet al., 2015)

ITER-like <lted plate
Fdiv,rad ~ 75%

What do we do with the Sectors: Blankets,Divertors, Launchers in the Hot Cells?
Inspect
Decontaminate (clean oﬀ)
Inspect
Dismantle
Inspect
Examine untreated surfaces
Examine mounts/connectors
Cut samples
FW
Side wall
Grid plates
Moun<ng hardware
SRDiv armor
Div structure
FCI
W stabilizer…...
Material examinaHons (PIE, mech prop tests, He bubbles, etc.)
Also examine the test specimens in the material test module

From ARIES-AT

The Hot Cell – a cri<cal aspect of the FNSF
The performance of materials in the components of the fusion core is not accessible
prior to the FNSF
What is accessible: 1) DONES/IFMIF or similar fusion relevant neutron exposure,2) ﬁssion neutron, ion and doping approaches tes<ng, 3) non-nuclear highly
integrated component tes<ng…..would we proceed if #1 was missing?
The in-service condi<ons include more than neutrons and temperature (like IFMIF)
Materials are under stress (pressure)
Materials have hydrogen in their matrix, tri<um and transmuta<on hydrogen
Materials have contact with liquid metal in B-ﬁeld (MHD, other ﬂow proper<es,chemical reac<ons)
Materials see strong neutron damage, and He produc<on gradients into material
Materials see hea<ng, temperature and stress gradients
The sectors of the fusion core will be removed to determine property changes, surfacechanges, microstructure changes.....this is done in the Hot Cell, and will produce thedatabase upon which DEMO can be based
we need to make decisions about the materials/components behavior for next phases
based on this informa.on in the FNSF itself…..turnaround must be fast, materials willbe HOT (radioac.ve)

Temperature. Deg C

The Complex Varia<ons of theService Environment in a Blanket
Helium production (appm) for
100 dpa at plasma facing sideplasma
F82H Structure(Max(T) ~ 530 ℃)

LiPb, ~3MPa

15001000500 250

530 C

400 C

DCLL Blanket
plasmaPLASMAappmHe100 8060 40

20He, 8 MPa

Von Mises stress

dpa
50mm
25mm
0mm
H. Tanigawa, E.Wakai 2012

Y. Huang, N. Ghoniem, UCLA

0.6 MPa

94 MPaPre-FNSF: Fusion Nuclear Materials Science, how do we see providing
tested materials in the form of components to the FNSF
20202030Non-nuclear characteriza<on
Fission, ion and doping irradia<ons
Fusion relevant neutron irradia<ons
Industrial/manufacturing
Material/environment match

2050

2040
DD

2060

DT FNSF
7 dpa

19 dpa26 dpaPre FNSF RAFM-1 development
Pre FNSF RAFM-2 development
Pre FNSF RAFM-3 development
Pre FNSF RAFM-4 develop
Pre FNSF FCI/SiC-c-1 development
Pre FNSF FCI/SiC-c-1 development
Pre FNSF baini<c development
Pre FNSF tungsten-1 development
Pre FNSF tungsten-2 development

US DEMO
37 dpa37-74 dpa

# of samples of mech type# temperatures# materials
Test volDpa/FPYAvailability

à What type ofdatabase is required forFNSF? ScienHﬁc orengineering?
A quan<ta<ve analysis of a single
blanket concept could make theurgency case for ge}ng to fusion
relevant neutrons NOWStar<ng point for organizing the pre-FNSF: PFC/
PMI area from the FNSF perspec<ve
SOL plasma experiments/diagnos<cstokamaks SOL/divertor plasma simula<ons
Divertor plasma experiments/core coupling
Plasma science

Establish “transla<on” oflinear results to tokamaks

Plasma material interac<on/linear/ultra-long dura<onPlasma material interac<ons/tokamak/short pulse
PMI science
Loading condiHons

Materials science and PFC component development
High heat ﬂux simulators (ebeams, lasers, ﬂashlamps, etc.)

TesHng of PFC candidates inlinear and tokamak faciliHes

Liquid metal test stands, plasma/vacuum, LM proper<es, ﬂow, geometry, LM species,
substrate design and fabrica<on, etc
Solid materials science, PFC component design and fabrica<onThe FNSF is the Cri<cal First Step into the Fusion
Nuclear Regime and Fusion Power Produc<on
The complex step of combining the fusion neutron,mul<-factor environment, and integrated components
requires the appropriate technical pla~orm à thisdemands a smaller intermediate device step
Explore the component material responses to thecomplete/correct FUSION environment
Establish the ultra-long plasma pulse length withassociated plasma material interac<ons
Establish the credibility of genera<ng tri<um, predic<ng/controlling its movement
and reten<on, and minimizing its losses plantwide
Strongly advancing most features toward a fusion power plant, with signiﬁcant fusion
relevance imposed in our technical choices
Understand the demands on the facility (hot cell, maintenance, diagnos<cs) to provide
the needed measurement, inspec<on, access, and feedback on plasma and fusion core
opera<ons

TUESDAY 1:15 PM
L. El-Guebaly, Overview of Nuclear Analysis for US Next Step Facility
J. Blanchard, Eﬀects of ELMs and Disrup<ons on FNSF Plasma Facing Components
L. Garrison, The Inﬂuence of Microstructure on Deuterium Reten<on in Polycrystalline
Tungsten…poster Monday
Poster Tuesday, Y. Zhai, Modeling Imperfect Nb3Sn SC Wire Under Transverse Loading
WEDNESDAY 8:50 AM
P. Titus, IN-Plane and Out-of-Plane TF Coil Support for the US FNSF Reactor
A. Davis, Overview of State-of-the-Art Neutronics Code Development…poster Monday
WEDNESDAY 10:30 AM
P. Humrickhouse, Safety in the Fusion Nuclear Science Facility
THURSDAY 10:15 AM
Y. Katoh, Progress in US/JA PHENIX Project for the Technical Assessment of PlasmaFacing Components for DEMO Reactor