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Russian Federation

DEMO program

(Development Status of the Russian DEMO Project )

Boris Kuteev

[email protected]

IAEA DEMO-6, Rosatom Tech, Moscow, RF, October 1-4, 2019

NATIONAL RESEARCH CENTER

“KURCHATOV INSTITUTE”

⚫ The Russian pathway to the Controlled Fusion (CF) with

magnetic confinement is inseparably linked with the

tokamak device and its enabling technology development

⚫ Demonstration experiments at the GW-level of fusion

power have been the dream of RF-scientists since

Kurchatov era

⚫ Those declared the neutron production rate of “one gram

per day” as the short term goal in fifties of previous century.

This rate corresponds to the DT-fusion power of 6,4 MW

that is close to the power of the first nuclear power plant

built that time in Obninsk.

⚫ The DEMO scale facilities of GW fusion power had passed

through the pre-conceptual design twice

in 70-ties (T-20, V.V. Orlov et al.) and

at the edge of the last millennium

(RF DEMO-S, Yu.A. Sokolov & G.E. Shatalov et al.).



RF DEMO-S layoutMajor radius, m 7,8

Minor radius, m 1,5

Elongation 1,85

Plasma current, MA 10

Toroidal field, T 7,72

Pulse duration, day 1-10

Fusion power, GW 2,44

Thermal power , GW

- ceramic blanket

-lithium cooled blanket

~ 3,1

~ 2,6

Blanket thermal power, GW

- ceramic blanket


~ 2,8

~ 2,3

Electric power, MW

- gross

- net

1100-1200

600-700

Neutron wall loading, MW/m2

- average

- maximal

2,52

3,4

NBI heating power, MW 100 - 120

Life-time, year

- throwaway member

- permanent set

~ 8

~ 20

First wall fluence , Mwy/m2

- ceramic blanket


-permanent set

~10

~10 - 16

~50

Pulse number

-blanket/divertor

-permanent set

~1500 /150

~7000 /700

First wall loading, MW/m2

-average

-maximal

0.4

0.7

Total disruptions 100

2002 DEMO-RF design

Yu.A. Sokolov, G.E. Shatalov et al.

DEMO-S blankets

Lithium cooled

blanket

Ceramic helium

cooled blanket



DEMO-S design findings

⚫ Operation is restricted by volumetric and surface heat loads of in-

vessel elements

⚫ The design requirements for the first wall heat fluxes 0.7 MW/m2 and

divertor targets 10 MW/m2 and neutron loading 3.5 MW/m2 are

affected by:

⚫ the limited life-time of the first wall due to erosion (at a chosen thickness)

⚫ necessity to operate in steady state regimes with a limited number of

disruptions

⚫ necessity to provide tritium breading ratio higher than 1.05 during the

lifetime period of the first wall and blanket for the design accepted

⚫ maximal temperature and thermal stresses in divertor targets



⚫ After starting the ITER project in 2006 the DEMO activity in Russia

practically stopped. Meanwhile, Russia had involved actively in realization

of ITER project hoping to use the ITER results in realization of CF based

on tokamaks.

⚫ Accounting for the growing international activity within ITER project, a new

pre-conceptual design of the Russian DEMO plant has been started this

year in NRC Kurchatov institute. The construction of this facility is

foreseen after 2055.

⚫ The major goals of this DEMO-design and corresponding R&D activity

include

-selection of basic fusion technologies appropriate for tokamaks with up to

the GW level of electric power and

-providing the long-term operation at DT-fusion power up to 40 MW as the

first step.

⚫ The principal device of this direction will be divertor tokamak T-15MD with

copper coils that will reach full scale operation with heating power up to 20

MW by 2024.



Milestones of the Rosatom Fusion Strategy (2007)

2005 2010 2015 2020 2025 2030 2035 2040 2045 2050

Burning

plasma

demonstration

Fusion plasma

physics

Materials

and

technologies

Facilities,

reactors,

power plantsТ-15D

ITER

FNS-1

DEMO-PP PROTO-PP

ITER

Q=5

400 s

ITER

Q =10

3000 s

Transport and stability

Т-10, Globus-М, Т-11М

Q~1

Physics in

Т-15D

ITER

superconductors

ITER

equipment,

DEMO TBM

Tritium

technology

demonstration

DEMO

Long term operation at Q~30

Q =30

Physics in

DEMO

New materials and

technologies for

5 MW/m2

Technical

requirements

DEMO-PP

Q =10

Physics in

ITER

DEMO

equipment

FNS-2Fusion-Fission Hybrid Systems

FNS-0

⚫ A specific feature of the Russian way to CF associates with the

interest in fusion-fission hybrid systems, which produce and use

fusion neutrons for control of subcritical fission active cores.

⚫ This line of development continues in the contemporary CF

research. Development of Fusion-Fission Hybrid Systems (FFHS)

forms the second direction of the program that supports both

fusion development and innovative fission technologies needed for

enforcement of the nuclear industry in Russia.

⚫ The hybrid reactor facility based on superconducting tokamak with

fusion power 40 MW and fission power 400 MW DEMO-FNS

should be designed and constructed by 2033. It will demonstrate

hybrid technologies for fuel nuclides production for thermal and

fast fission reactors as well as transmutation of minor actinides.



⚫ FFHS will accelerate fusion development via demonstration of

SSO and remote handling tokamak technologies as well as

industrial tritium production for the first loading of fusion power

plants.

⚫ The program must upgrade the currently available

technologies including low and high temperature superconductors,

plasma heating and current drive technique and others up to the

pilot level of manufacturing that is needed for construction of the

program facilities.

⚫ Successful realization of the technical part of the program

requires a substantial involvement of RF nuclear regulator

(Rostechnadzor), licensing and general design organizations. The

future goals beyond 2035 include the Pilot Hybrid Facility (2045),

Fusion Power Plant (scale-beyond 2055) and Commercial Hybrid

Plant (2055).



(In-kind Contribution to ITER)

TF Coils (18)

Feeders (31)

CC Coils (18)

PF Coils (6)

CS Coils (6) Divertor

Blanket

Vacuum Vessel

Thermal Shield

Cryostat

Feeders (31)

• The largest international RF project based on in –kind cooperation

• EPC (Engineering, Procurement and Construction) project with R&D part

• Compatibility of systems produced by different Suppliers

• Systems must satisfy requirements of Nuclear Regulator in France (Installation Nucléaire

de Base - INB)

ITER-RU

TBM in ITER

Ceramic helium cooled TBM

Lithium cooled TBM

TBM RF



Majorradius R, m 1.48

Aspect ratio 2.2

Plasma current IP, MA 2.0

Elongation 1.9

Single and double null divertor

configurations

SN,D

N

Pulse duration, sec 10-30

Toroidal field, T 2

Flux stored in the central solenoid, Web 6

Neutral beams power, MW 6

Gyrotrons power, MW 7

Ion cyclotronheating power, MW 6

Lowhybrid heating power, MW 4

Basic parametersDivertor tokamak Т-15MD

Major RF facility supporting

ITER activities in RF

Full power operation at 2024

Heating and current drive, Divertor,

Lithium technologies, Diagnostics

NATIONAL RESEARCH CENTRE

KURCHATOV INSTITUTE

НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ

ЦЕНТР «КУРЧАТОВСКИЙ ИНСТИТУТ»


KURCHATOV INSTITUTE

НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР

«КУРЧАТОВСКИЙ ИНСТИТУТ»

T-15MD construction status (June 2019)

• Toroidal

coils,

vacuum

vessel and

central

solenoid are

assembled

• Vacuum

tests have

begun

NATIONAL RESEARCH CENTRE

KURCHATOV INSTITUTE



Energy valuable system needs a combination of 6 parameters

• n20 – plasma density in 1020 m-3

• TkeV – plasma temperature in keV

• tE – energy confinement time in s

• kg – Kurchatov neutron yield in g/day

• tSS – steady state operation time in y

• C – duty factor

Breakeven/Ignition

Q = PFusion/PAH

proportional to

the Triple product n20 TkeV tE

Controlled Fusion/Energy

Kg = n20 TkeV tE kg tSS C

Facility n20 TkeV tE kg tSS C Q Kg

JET 1 10 0.3 0.35 3.5x10-7 0.1 1 3x10-8

NIF 1012 0.2 2x10-11 10-8 10-6 0.1 0.015 4x10-15

ITER 1 10 3.5 25 10-4 0.25 10 2x 10-2

FNS-ST 1 2 0.05 0.2 1 0.3 0.2 6x10-3

DEMO-FNS 1 4 0.3 2 1 0.3 1 7x10-1

DEMO 1 15 5 50 1 0.5 25 2x103

PROTO 1 15 6 150 1 0.8 30 1x104

Transition from Modern Tokamaks to PROTO -> 12 orders growth of Kurchatov factor Kg

Synergy of Fusion and Fission


KURCHATOV INSTITUTE



Only FFHS are potentially neutron rich

Factors providing unique efficiency of FFHS in free neutron generation:

Multiplication of ДТ- neutrons

in reactions Li, Be, Pb(n, 2n) (n,3n) и fission HM (n, nn) FP

Growth of neutron yield (linearly) in fission reactions

with neutron energy for (238U) 3(Fission-n) ->

4.5 (Fus-n)

G-factor 14 MeV ~1.5 for first generation neutrons in 238U

with mass of HM 4,5 (U) -> 6 (МА)

G-factor 14 MeV ~2 - first generation

spectral additives >2.7 - 1-5-th

generation

Subcritical fission of HM

G x keff

multiplication factor M = 1- keff

Additional source

on delayed neutrons adds ~ 0,023/0.05

up to~5 % in power and neutrons

Free Neutrons in Fission and Fusion Systems


KURCHATOV INSTITUTE



SYSTEM

Neutrons

per fission n ,

multiplication m

Residue

(reaction

maintenance and

breeding

subtracted)

Leakage and

volumetric

losses

Free neutrons

available for

extended

breeding

Critical nuclear

reactor 235U

thermal spectrum

n

2.44

n - 2

0.44 1.0 -0.56

Critical nuclear

reactor 239Pu

thermal /fast

spectrum

n

9. / 3.05

n - 2

0.9 - 1.05 0.9 0 – 0.15

Fusion reactorD+T

Be /Pb - multiplier

m

2. / 1.8

m - 1

1 / 0.8 0.8 0.2 / 0

Hybrid reactor

D+T and 238U

n n - 1

3.5 1.0 2.5

Hybrid reactor

D+T and MA

G=2 keff = 0.95

keff = 0.8

Gkeff /(1-keff )

38

8

m - m/(nG)

38 - 6.2 = 31.8

8 - 1.31 = 6.69

12.5

2.64

19.3

4.05

Hybrid reactor (MA) - the best option for T-loading and maintenance of Fusion reactors

Highest multiplicationtritium

breeding

impossible

marginal

marginal

Breeding:

reasonable

Fusion Strategies

RF 2033 2063

-----------------------------------------HRF design+15 y R&D+9 y construction HRF-Conventional Tokamak (~30 y)

licensing2055

---------------------------------------------------------------------------------------DEMO-RF concept, CDA, EDA, licensing, construction DEMO-RF ОТЭ

--------------------? CFPP ПТЭ

------------------------FNS design, R&D, construction

Fusion neutron source is needed for materials and components development and

attestation

Compact FNS develop within IAEA CRPs since 2012

US Final

Report 2018

Major facilities on the path to Commercial Hybrid Plant

• Magnetic system• Vacuum chamber• Divertor• Blanket• Remote handling• Heating and current

drive• Fuelling and

pumping• Diadnostics• Safety• Molten salts

Pilot Hybrid Plant construction by 2045 P=500 MWth, Qeng ~1

Steady State Technologies

•Materials

SSO&MS Globus-3 FNS-ST DEMO-FNS

DT neutrons MS blankets

•Hybrid Tech•Integration

центральный столб обмотки тороидального магнитного поля вакуумная камера плазменный шнур опорная структура

Investment $1 B $0.1 B $1 B $5 B

$10 B

$100 B Commercial Hybrid Plant construction by 2055 P=3 GWth, Qeng ~6.5 P=1.3 GWe, P=1.1 GWn, MA=1ton/y, FN=2 ton/y

T=25 kg /y. Fusion /Fission Power = ~ 1 %!! 1 CFPP start/y

Last design activity confirms the time and cost scales

FP Task 2 Roadmap

Fusion Energy and Nuclear Fuel Cycle

Nuclear Power

Applications

Hybrid technology demonstration

Attestation of materials and components in Fusion neutronsIntegration of tokamakSSO technologiesTestbeds for SSO

enabling

technologies

Design of Testbeds

CHP40+3000 MW,

PPP/DEMO 1 GW

PHF 40+400 MW

Electro-tech. Neutron PHF (eng. des.) Plasmatechnologies technologies Т, N-Fuel, trans., En., DT-n technologies

DEMO-FNS 40 + up to 400 MW

FNS-ST 3MW

Globus-3

Tokamak Tritium Remote H. Blanket Hybrid NFC Molten salts

2055

2045

2035

2033

2026

2023

2022

2019

DT-fusion + fission

Research and development of hybrid reactor technologies and systems

FNS-ST: basic parameters and cut-view

R, m 0.5

R/a 1.66

k 2.75

δ 0.5

Ip, MA 1.5

Bt, T 1.5

n, 1020m-3 1 - 4

Pwall, MW/m2 0.2

Eb, keV 130

Pb, MW 10

Angle NBI, deg 30

PEC, MW 5

H-factor 1-2

βN 5

fnon-ind 1.0

Pdiss, TF, MW 14

Pdiss, PF, MW 6.0

Swall, m2 13

Vpl, m3 2.5

Roadmap for FNS-ST design and construction

2009 2010 2011 2012 2019 2020 2021 2022 2023 2024 2025

Data collection

Draft proposal

Preliminary specifications

Conceptual design

Engineering design

Detailed design

Building and site

Power supply

Water supply system

Assembling tools

Cooling system

Cryogenics

Magnets

Vacuum chamber

Divertor

Blanket

Heating and current drive

Fueling and pumping

Diagnostics

Emergency system

Safety

Control and data acquisition

Licensing

Remote handling

Radiochemistry

Tokamak based

fusion neutron

source for

testing

materials and

components

in NRC KI-PINP,

Gatchina

Fusion neutron

source

1- FNS-ST Building

2-Assembling Hall, Gyrotron Hall

3-Hot Cells, Spent Fuel Depository, Tritium and Pumping Systems

4-Chemical Processing, Hot Cells, Depository of

the Fuel Breeded

5-Low Active Waste Hall

6-Gas Storage

7-FNS Access Hall

8-Diagnostics, Capacitors, Reactors Hall

9-PowerSupplies for TFC

10-Magnetic System Transducers

11-NBI Power Supplies

12-Cryoplant

13-Pulsed Power Supplies Area

14-Emergensy Power Supply

15-High Voltage Transformers Area

16-Control Building and Laboratories

17-Pumping Station

18-Water Cooling Station

Total Area 130 m x 200 m =25200 m2


KURCHATOV INSTITUTE



Tokamak ST-40, Tokamak Energy, Great Britain

(M. Gryaznevich, AAPP Nature conference, SPb, RF 18-20.09.2019)

This tokamak started its 2nd operation campaign with neutral beam injection, exactly

according to the time schedule and budget proposed in our conceptual design

report 2012

Proposed accommodation on the site

Option placement of FNS-ST facility on the site of NRC KI-PINP in

Gatchina, Leningrad region. The site is adjacent to the PIK research

reactor complex, which provides research using neutron scattering



Research Reactor PIK, PINP, Gatchina, Leningrad region, 30 km from SPb

Start of operation 2019 (10 MW)

Power 100 MW, neutron yield ~1018 n/s, thermal flux 1015 n/cm2s



Status of DEMO-FNS design


KURCHATOV INSTITUTE



Aspect ratio R/a, 3.2m/1m

Toroidal Field 5 T

Electron/ion

Temperature, 11.5/10.7 keV

Normalized βN 2.1-2.4

Current Ipl 5 MA

Neutron yield GN 1.3·1019/s

NBI power 36 MW

ECRH 6 MW

Discharge duration 5000 h

Duty factor 0.3

Life time 30 year

Consumed/

Generated power 200/200 MW(e)

Thermal power 500 MW (th)

Total thermal power 700 МВт т

Tritium on site 2 kg

T-consumption/breeding 700 g/year

Fissile nuclides breeding 48 kg/year

• The facility is considered as the major source of fusion scientific and technology

information supplementing ITER project in development of Fusion Fission Hybrid

Systems and Fusion Power Plant

MAactinides 1-20 ton

Cost ~ 3 $B

Site area 50 h

Staff 1200


KURCHATOV INSTITUTE



2018 design of D.V. Efremov Institute, SPb

Generalview of DEMO-FNS1. Top part of the biological shield

2. Cryostat part of the biological shield

3. Cask transporter

4. Removable plug for the top shield

5. Circular–action loading and unloading machine

6. Galery for bass bars and cryogenic piping

7. Galery for systems of additional heating and current

drive

8. Galery for water coolant piping of the vacuum vessel

Upper port

Passive

loops

Vacuum

vessel

NBI injection

portFirst wall

VVsupport

Pumping port

Pumping system

Blanket

Equatorial port

Divertor

Cryostat


KURCHATOV INSTITUTE



Neutral Beam Injection (Kurchatov & Budker)

Schematic of NBI injectors siting around the vacuum vessel in the Machine hall of DEMO-FNS: 1 — poloidal coils; 2 — toroidal coils; 3 —NBI guide line; 4 — port of NBI-injector; 5 — equatorial port; 6 —vacuum vessel; 7 — first wall; 8 — plasma; 9 — criostat; 10 — shield 1,25 m (—) and2,0 m 11 — gate valve, shutter; 12 — beam collector, neutrolizer; 13 — ion source

Schematics of NBI injectors. а — gas mixture with equal fractions of D and T; б — Deuterium enriched gas mixture

D+T

DMachine hall

50x50 m2

DEMO-FNS integration, blanket and remote handling


KURCHATOV INSTITUTE



Configuration of active

cores and tritium breeding

blanket

Subcritical

active core with

Minor Actinides,

30 MW,

2х0.8х0.15 m3

Conceptual design of

Remote Handling

System

JSCs NIKIET and

NIIEFA

DEMO-FNS divertor


KURCHATOV INSTITUTE



Schematic of 3-chamber divertor

DEMO-FNS.

A source of Lithium vapor jet is

placed in the upper part of box 3.

Jet reduces the lithium flux from box

2 in box 3 and SOL

Lithium technologies using

vapor, liquid, dust and thin foils

interacting with plasma and surface

of in-vessel components are

considered as prospective technical

solutions providing long life in-

vessel components

Technology was partially tested on T-10

will be further developed and tested on T-

15MD, FNS-ST and DEMO-FNS.


KURCHATOV INSTITUTE



Kikuchi slide demonstrates

operation domains of FNS-ST and DEMO-FNS

DEMO-

FNSFNS-ST

DEMO-

FNS

~timp-0.5

~timp-0.3

~timp-1.0

Heat conductivity

Heat capacity

Heat transfer

Discharge Duration Limits of Contemporary Tokamaks and Stellarators


KURCHATOV INSTITUTE



•Duration limit

may be a product of

local heating with wet

area less than 0.5 m2

•No-coolant operation

is affected by radiation

•Coolant rate

of 10th MW/m2 may not

help against hot spots

Critical temperature

•Limitation of the pulse duration typical for contemporary MF-devices is a serious obstacle on the way to fusion energy applications.•Plasma-surface interaction and heating of plasma-facing components to high temperatures remains the major factor. •Experimental data available (TFTR,JET, LHD, DIII-D, Tore Supra, EAST, NSTX, KSTAR) can be explained on the basis of heat exchange of SOL plasma with the localized surfaces of the device, taking into accounts coolants and FW-radiation.•The wetted area responsible for the pulse termination is significantly smaller (≤1%) than FW-area. •Local heat fluxes higher than 70 MW/m2 practically exclude SSO for contemporary technologies. •Further development of fusion reactors requires more uniform FW-heat loads and closer attention.

no coolantcoolant

(BK&VS NF submitted)

no coolant

coolant

coolant

no coolant

Critical loading and exhaust

Experiments& Simulations merged bySwet

P/S, P/R parameters

are inapplicable!

• Advanced confinement of burning plasma

• Steady state heating and current drive

• Control of plasma stability

• Equilibrium and shape control

• Extended operational limits (НУ2, βN, IN q95, к, δ, n/nGr)

• Plasma-wall interaction defining the operation life (impurity control,

materials, thermal, neutron and fast particle loading, erosion,

redeposition, dust, recycling, permissivity etc.)

• Fueling and particle flows control in steady state, optimal divertors

• Diagnostics compatible with neutron environment and SSO

• Reactor neutronics and blanket physics

• Tritium breeding and heat transfer in blankets

• Physics of hybrid blankets

• Development of databases for fusion physics, nuclear physics and

materials properties

Strategy tasks in Physics



DEMO challenges⚫ Physics and engineering

stability, self-organized plasma, fast particles, disruption,

non-inductive discharges

⚫ Steady state operation technologies

first wall, heating & current drive, fueling & pumping,

diagnostics, control

⚫ Materials structural, functional, coolants

⚫ Tritium circulation level of kg per hour

⚫ Remote handling in

SSO environment 0.2-2 MW/m2 14 MeV neutron loading

⚫ Tritium breeding with 1-100 kg/year rate

⚫ First wall and divertor with liquid metal protection

⚫ Integration

⚫ Regulations

⚫ Staff



NRC KI DEMO teamKurchatov Complex for

Fusion Energy and Plasma Technology

⚫ Physics and simulation Dr. Vladimir Pustovitov

⚫ Materials Dr. Alexander Sivak

⚫ Fueling and auxiliary heating Dr. Sergey Anan”ev

⚫ Diagnostics Dr. Alexander Melnikov

⚫ Design and Integration Dr. Yury Shpanski



Conclusions


KURCHATOV INSTITUTE



• Russia searches for a pathway to controlled fusion together with fission

power in frames of the SC Rosatom’s State Program

• The pure fusion activity is supported by projects of T-15MD and Globus-M2

as well as by participation in ITER (definitely effective) and recently started

domestic DEMO oriented R&Ds and pre-conceptual design

• Fusion-Fission Hybrids are considered in Russia as an important player in

the global nuclear energy development. We believe that FFHS are needed

both to Fusion and Fission Power engineering.

• Upgraded Fusion Program has been proposed in RF and it circulates

waiting an approval from federal jurisdictions (2019 start still possible)

• The Program Tasks will develop M&I fusion, FFHs, FF enabling

technologies and their applications, FNS, nuclear regulation, staff

education and training

• Development of Fusion and Fusion-Fission Hybrid Systems requires

design and construction of facilities, materials and technologies of new

generation which will be activated by the Program approved

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