Upload
others
View
7
Download
0
Embed Size (px)
Citation preview
Russian Federation
DEMO program
(Development Status of the Russian DEMO Project )
Boris Kuteev
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.).
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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
-lithium cooled 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
-lithium cooled 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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
⚫ 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.
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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.
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
⚫ 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).
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
(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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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»
NATIONAL RESEARCH CENTER
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
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
Status of DEMO-FNS design
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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.
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
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
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
•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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
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
NATIONAL RESEARCH CENTER
“KURCHATOV INSTITUTE”
Conclusions
NATIONAL RESEARCH CENTER
KURCHATOV INSTITUTE
НАЦИОНАЛЬНЫЙ ИССЛЕДОВАТЕЛЬСКИЙ ЦЕНТР
«КУРЧАТОВСКИЙ ИНСТИТУТ»
• 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