1National Research Center ‘Kurchatov Institute’, Moscow, Russia 2State Polytechnic University, Saint Petersburg, Russia

3Tokamak Solutions, Abingdon, UK

Developing FNS-ST design codes A.Yu.Dnestrovskij1, B.V.Kuteev1, A.Golikov1, S. Yu. Medvedev1, A.A.Ivanov1, V.E.Lukash1, R.R.Khayrutdinov1, A. Zhirkin1, A. Sivak1, A.Spitsyn1,

S. Anan'ev1, M.I.Gurevich1, V. Yu. Sergeev2, A.S.Bykov2, P.R.Goncharov2, M.P.Gryaznevich3,

2nd IAEA DEMO PROGRAMME WORKSHOP

IAEA Headquarters, Vienna, Austria

17-20 December 2013

Abstract

Design tools for fusion neutron sources implemented for FNS-ST project [B.V. Kuteev et al., Nucl.

Fusion 51 (2011) 073013] are reviewed. Indefinite physical parameters playing a key role on the

project are considered here. The ranges of their change are defined by codes, calculations and

experiment. The sensitivity of the calculation results to these parameters is determined by the scanning

parameter values within the reasonable range. The basic plasma parameters for FNS-ST device are the

following:

DINA 1D free boundary, time evolution

1-10

MCU Neutronics, 11,12

Material code (MD), 11

Tritium cycle, 6,11,13 SANYC semianalytical

treatment for fast particles 4,7,10

SOL-Onion-skin

Divertor plasma 9,11

SPIDER Free boundary equilibrium

1,8,11

KINX RWM instability

1,8,11

FNS-ST design tasks 1.Geometry (R,a,k,δ) 2.Scenario (ISOL, IPF(1-8), IPL) 3.Powers (PNB,ENB,PEC) 4.Plasma current (BS,NBCD) 5.Confinement 6.Particle transport 7.Fast particles 8.Profiles (pedestal, j, li) 9.Power load (Rad, Div) 10.Fusion 11.Construction 12.Neutronics 13.Tritium cycle

Parameter, FNS-ST values Defined by

n<nGr~5∙1020m-3 Greenwald limit

βN <6 Stability (KINX)

H-factor~1-1.4 Experiment, variation

Geometry: δ, k, q95, shaping factor =INq95 ~ 1/A(1+k2(1+2δ2)) ~30

Stability (KINX)

Poloidal flux consumption DINA (Bv and Central Solenoid ramp up)

Profile shape, pedestal, internal inductance

Experiment, Stability (KINX), Control requirements (DINA,SPIDER, γ<100s-1 )

Beam-plasma fusion 0D, SANYC, NUBEAM

Current drive efficiency, ηCD =RICD<n>/Pb

NUBEAM ηCD =0.12A/m2/W, Experiment ηCD ~0.1A/m2/W

Beam energy, Eb=130keV Technology,

Radiation ~20% main plasma, ~10% SOL ASTRA,DINA,SOL

Fast particle losses, τfastNBI > τslowdown , τα ~ 0 Experiment, NUBEAM

Parameter, FNS-ST values Defined by

Plasma density at separatrix, nu=1/3<n> Experiment on AUG

Power fraction to the divertor plates ~50% Input power = main plasma radiation + heat flux on wall + fast particle losses + divertor plates heat flux

Experiment

Heat flux asymmetry for inner – outer divertor plates 1:4 single-null and 1:9 double null divertors

Experiment

Expansion factor for heat channel

(Divertor plates angle and expansion of magnetic field surfaces)

Geometry and DINA calculations

SOL width, λq~0.8cm Experimental scaling law

Impurity fraction External parameter

Patameter Value Dim

Vacuum vessel volume 2,50 m3

Vacuum vessel wall area 13,00 m2

Plasma density 5E+19 m-3

Fusion power 3 MWt

Particles confinement time

τparticle ~ (1-4) τE 50 ms

NBI injectors number 4 psc

NBI power 10 MW

NBI energy 130 keV

k(nbi) 100 %

k(pellet) 90 %

k(puff) 20 %

Gas flow to the hydrogen

isotopes separation system 2 %

Main plasma key parameters

SOL plasma key parameters

ASTRA-NUBEAM

0,5

1,0

1,5

Bad orbit losses, MW

Ipl,MAFNS-ST

0,6 0,8 1,0 1,2 1,40

2

4

6

N

Neutrons,1017

/cH-factor

0,0 0,1 0,2 0,30

10

20

<n>=1020

m-3

Density, 1019m-3

minor radius, m

0,00

0,05

0,10

0,15

Stot

=10 1017/s

Stot

=9.5 1017/s

-40 -20 0 20 40 600,0

0,2

0,4

0,6

0,8

1,0

Bootstrap

NBCD

Cu

rre

nt,

MA

Vertical beam angle, deg

Beam injection angle variation with NUBEAM code

Density profile variation – no strong effect on fusion power

NSTX FNS-ST

Ipl, MA 0,6 1

ψres, Wb 0,28 0,09

ψli, Wb 0,19 0,11

LeIpl, Wb 0,24 0,37

∆ψCS, Wb -0,51 -0,28

∆ψBv, Wb -0,2 -0,29

DINA Free boundary equilibrium current ramp up scenario

Bv ramp up flux ~ Central Solenoid flux 0 50 100 150 200 250 300

-150

-100

-50

0

50

100

150

r [cm]

z [cm

]

PF4

PF1 PF2

PF3 Ip, MA 1.5

q95 5.3

Bt, T 1.35

t 0.17

N 4.18

li(3) 0.43

ksep 2.87

a, cm 27.5

up 0.35

dw 0.35

Rmag, cm 54.8

Rsep, cm 38.0

Zsep, cm 79.0

-0.5 0 0.5 1 1.5 2 2.5

-1

-0.5

0

0.5

1

-0.5 0 0.5 1 1.5 2 2.5

-1

-0.5

0

0.5

1

coil currents [MA]

PF1 0.6358 PF2 1.8030 PF3 -0.0842 PF4 - 1.3042 CS -0.5000

coil currents [MA]

PF1 1.7067 PF2 1.5415 PF3 -1.3984 PF4 -0.6447 CS -0.5000

li =0.28 PF1

PF2

PF4

CS

PF3

PF3 - nearly optimal coil position for plasma shaping under profile variations

R/a 1.75

k 2.75

δ 0.48

R/a 1.75

k 2.72

δ 0.33

li =0.41

Calculation for steady state regime: Optimal PF coils position and currents

for different plasma current profiles (different li);

SPIDER Free boundary equilibrium and evolution code

ASTRA-NUBEAM 1D fixed boundary, steady-state

3-10

0D code 1,3,4,5,7,9,10

NSTX

KINX ideal MHD stability code

• External kink mode (toroidal mode

numbers n=1-5) limits for βN • Peeling-ballooning mode stability diagram

with high toroidal wave numbers n=3 RWM structure

(li =0.28)

• RWM

0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2

-1

-0.5

0

0.5

1

γ=601/s

0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2

-1

-0.5

0

0.5

1

VDE t=0.049s

Vertical n=0 instability growth rates.

0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2

-1

-0.5

0

0.5

1

The outer wall stabilization

Outer wall:

2mm CuCrZr

instead of

20mm steel (3 times

less resistance)

2cm CuCrZr

1. Current limit (β=0, n=1)

Maximal IN < 8 (3MA, B=1.35T)

Operational IN < 5 (1.8MA)

stationary with high β

2. β-limit

Global kink modes: βN < 6

RWM stabilized, rotation and kinetic

effects, q0 > 2

Medium n>5 kink/ballooning: βN < 6

3. Pedestal and ELMs

ELM regime?

v I, kA*t, DINA

I, kA*t, SPIDER

PF1 809 635

PF2 377 1800

PF3 - -84

PF4 -864 -1304

CS -500 -500

Inductive current ramp up

v

H-factor has strong effect on neutron output

Tritium cycle

No strong influence of τparticle on total tritium flow because less than 10% of it is going through the vacuum vassel

Fusion specifications:

High neutron energy

Threshold cross sections

MCU Neutron transport and nuclear transmutation

Thorium fuel cycle in FNS-ST

blanket: Eutectic salt FLiNaK

with 15% of Th232F4 solution

Neutron adsorption in

construction material:

• W - 2.77685E-01

• Cu - 2.17398E-01

• K - 1.34977E-01

• F - 2.43384E-02

Pb(n,2n)

Energy, MeV

Cro

ss s

ecti

on

, bar

ns

R R/a κ δ Ipl Bo N20 Ebeam Pbeam Fusion

0.5m 1.67 2.75 0.5 1.5MA 1.5T 1.1 150keV 6MW 1.7MW

Codes implemented for FNS-ST design

Total tritium accumulation

Total flow Flow through Vacuum Vessel

Tritium losses

burning decay

730g 90mg/s 8mg/s 40g/(MW*year) 60g/year

Conclusions

The FNS-ST project as a system of parameters is considered in this Report. It is impossible to create a

fully consistent project with the unified system code because project parameters are determined by

essentially different physics. Ranges of parameters change and their influence on the each component

of the project system can be defined by a different code. But many parameters such as H-factor,

profile shaping, impurity transport, SOL parameters, cannot be defined by codes; moreover the

experimental scaling laws are of limited validity because plasma and wall conditions lay far beyond the

existing experiment. However the effect of varying the parameters can provide an estimate of it

influence and hence provide some degree of reliability.

The list of key parameters is produced as a guide of project development for future experimental

validation as well as for code calculations

SOL radiation 0.8 MW

Outer divertor plates heat load 0.6MW/plate

Inner divertor plates heat load 0.05 MW/plate

Peak heat flux density on the outer plate 3.8 MW/m2

Plasma temperature near divertor plates 10-30 eV

Plasma density near divertor plates 3 1019 m-3

Upstream temperature 46 eV

Currents in the control coils in steady state regime

Temperature dependencies of

hydrogen isotopes diffusivities in

bulk iron crystal (molecular

dynamics).

These parameters would help to

optimize the tritium accumulation

in walls.

Experimental research is needed

for composite (alloy) material

SOL-Onion-skin

Expansion factor Heat load

NSTX

FNS-ST

Model uncertainty

Material code (MD) Material durability and hydrogen accumulation

The low flow through the hydrogen isotopes separation system because for all systems fuel has an equal concentration of Deuterium and Tritium components