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IAEA International Atomic Energy Agency
Overview of Fast Reactors Technology
and
IAEA Activities in Support of its Development
Stefano Monti ([email protected])
Team Leader - Fast Reactor Technology Development
Division of Nuclear Power
EC Collaborative Project SILER:
Training Course on Seismic Protection of Lead-Cooled Reactors
Verona, May 21-25, 2012
IAEA
IAEA Organizational Structure
IAEA
Outline
Why Fast Reactors
FRs: technologies, history, current status and future prospective
Key Elements for future Development and Deployment
General Framework of Safety of Fast Reactors
IAEA FR activities and deliverables and the role of the TWG-FR
3
IAEA
Why Fast Reactors
4
Almost all nuclear reactors under
operation are thermal reactor, in which a
moderator slows the energy of neutrons
down to the thermal energies
In thermal reactors the amount of fissile
material consumed is greater than the
one generated from fertile nuclides
Only a small fraction of the energy
potential of natural uranium is exploited
Uranium resources are under potential
stress if only U-LWRs are deployed.
The deployment of FRs with high
breeding gain and short doubling times
(CDT) can help.
Pressurized Heavy Water
Reactors9%
Light Water Graphite Reactors
4%
Fast Breeder Reactors
1%
Pressurized Water Reactors
59%
Gas Cooled Reactors
6%
Boiling Water Reactors
21%
Reactor Types in Use Worldwide, Jenuary 2004
IAEA
Why Fast Reactors:
Conversion Ratio vs. Breeding Ratio
5
Condition for breeding
η neutrons are generated by fission per neutron absorbed in a fissile isotope of these neutrons 1 is needed to maintain the chain reaction
some others (p) are lost by radiative capture or by leakage
Then the number of fissile nuclei produced per fissile nucleus destroyed is:
B =η −1− p
If B<1 the reactor simply “converts” isotopes and B is called Conversion Ratio
If B>1 the reactor “breeds” fissile nuclides and B is called Breeding Ratio
IAEA
Why Fast Reactors: Favorable neutron economy with
respect to thermal neutron spectrum reactors
6
η is significantly
higher in the case of
fission induced by fast
neutron than thermal
neutrons
Neutron yields for various fissile atoms
(Source: A. Waltar, A. Reynolds)
Reactor types Natural Uranium Uranium 235 Uranium 233 Plutonium 239
Thermal 1.34 2.04 2.26 2.06
Fast < 1 2.20 2.35 2.75
Neutron yields ‘’ in thermal and fast spectrum reactors
IAEA
Why Fast Reactors: the ratio fission/absorption is
also higher, and actinides are preferentially fissioned
7
Fissile isotopes are likely to
fission in both thermal and
fast spectrum
However, the fission fraction
is higher in fast spectrum
Moreover, significant (up to
50%) fission of fertile
isotopes in a fast spectrum
IAEA
Why Fast Reactors: Great flexibility thanks to excess
of neutrons and transmutation performances
8
As first discovered by Enrico Fermi in 1944, the nuclear characteristics of transuranics
(TRU) cross sections in a fast neutron spectrum allow a great FR flexibility:
Breed (i.e. Conversion Ratio CR>1) Sustainability
Burn (TRU or MA), i.e. CR<1 Transmutation for waste management
Breed (e.g. Pu) and burn (MA)
CR~1: Self-sustaining.
Wide coolant and fuel type choice according to the objective, e.g. short
Doubling Time DT: Na and dense (e.g. metal) fuels
Wide range of MA content and different Pu vectors or TRU compositions
Few credible competitors
IAEA
Why Fast Reactors: Resource preservation
9
The high breeding ratio, combined with the multi-recycling of the spent
fuel, allows fast reactors to fully utilize the energy potential of natural
uranium, assuring a potential energy supply for thousand of years
88
270
627*
2640
8100
18810
21120
64800
150480
1 10 100 1 000 10 000 100 000 1 000 000
Ide
nti
fied
res
ou
rce
sT
ota
l co
nve
nti
on
alre
sou
rce
s
Tota
l co
nve
nti
on
alre
sou
rce
s an
dp
ho
sph
ate
s
Pure fast reactor fuelcycle with recycling ofU and all actinides
Pure fast reactor fuelcycle with Pu recycling
Current fuel cycle (LWR,once-through)
Years
* - based on 22 Mtu of phosphates, 6306300 of identified resources, and 10 400 500 of undiscovered resources
FR with CR>>1
IAEA
Composition of Spent Nuclear Fuel (Standard PWR
33GWd/t, 10 yr. Cooling time)
10
1 tonne of SNF contains:
955.4 kg U
8,5 kg Pu
Minor Actinides (MAs)
0,5 kg 237Np
0,6 kg Am
0,02 kg Cm
Long-Lived fission Products
(LLFPs)
0,2 kg 129I
0,8 kg 99Tc
0,7 kg 93Zr
0,3 kg 135Cs
Short-Lived fission products
(SLFPs)
1 kg 137Cs
0,7 kg 90Sr
Stable Isotopes
10,1 kg Lanthanides
21,8 kg other stable
Most of the hazard stems from Pu, MA and some LLFP when
released into the environment, and their disposal requires
isolation in stable deep geological formations.
Decay heat in a repository also determined by Pu and MA
Uranium,
95.5%
Plutonium,
0.9%
Other long
Lived Fission
Products,
0.1%
Longlived
I and Tc, 0.1%
Short-lived Cs
and Sr, 0.2%
Minor
Actinides,
0.1%
Stable Fission
Products,
3.1%
IAEA
Why Fast Reactors: Actinides’ (Pu, Am, Np, Cm) Burning
11
Radioactive waste decay time Storage space requirements
Recycle of all actinides in fast reactors provides a significant reduction in
amount, heat load and time required for radiotoxicity to decrease to that
of the natural uranium ore used for the LWR fuel (from 250,000 years
down to about 400 years)
IAEA
Why Accelerator Driven Systems
12
Fast neutron systems should be privileged for transmutation.
In case of Minor Actinide dominated fuels and no U (“U-free
fuels”), the effective delayed neutron fraction can be
unacceptably low, with consequences on the safety case, if the
core is critical.
A way out: a sub-critical, source driven reactor (ADS).
However, there is the choice between critical (CR<1) or sub-
critical cores (ADS).
IAEA
Fast Neutron Systems in a Closed Fuel Cycle:
Towards a more sustainable nuclear energy
Tomorrow’s generation of reactors / Fast Neutron Reactors:
The above and….
Multiplication by a factor 50 to 100 the
energy produced by a given amount of
uranium,
Minimization of volume, thermal load
and radiotoxicity of waste …
“Closing the fuel cycle”:
enough resources for thousands of years
Today’s generation of reactors:
Safe, reliable and competitive
Availability of secure resources (about 100 y at the present rate of consumpition)
Reprocessing of spent fuel for enhanced use of resources
Technical solution for waste management
IAEA
Why Fast Reactors…. are not deployed as LWRs ?
14
There is the the possibility of designing nuclear reactors without neutron
moderator, in which the chain reaction is sustained by high energy
neutrons (fast neutrons)
Fast reactors can convert Uranium-238 into Plutonium-239 at a rate faster
than they consume the fissile nuclides, allowing a full utilization of natural
Uranium which is composed mainly by fertile nuclides (the amount of
U235 in natural Uranium is only the 0.7%)
This potentiality of fast reactors has been recognized since the beginning
of nuclear power era, but the intrinsic characteristics of fast spectrum
require more complex and expensive technologies (both for the reactor
and the associated fuel cycle) which have prevented so far the same
successful development/deployment of light water reactors
IAEA 15
Fast Reactors:
Technologies, History, Currents Status and
Future Developments
IAEA
Three main technology options depending
on primary coolant
Sodium Fast reactor
Lead Fast Reactor Gas Fast Reactor
IAEA
Sodium Fast Reactor (SFR) outlook
• Strong national programs and
experience (400 reactor.year)
• Gather fresh operating
experience from existing, new
and restarting reactors
• Key technical issues Advanced fuels including actinide recycling
Converge safety approach (main issue: Na chemical reactivity)
Resolve feasibility issues regarding in-service inspection and
repair
Energy conversion systems
Codes and standards for high temperature application (550 C)
4
IAEA
Sodium properties: several advantages
Low melting point (97.8 C)
Large range of the liquid phase (99 C - 880 C)
Low density and viscosity
Very high thermal conductivity
Excellent electrical conductivity
Low saturation vapor pressure
Transparent to neutron
Low activation
No specific toxicity
Cheap and largely available
Perfectly compatible with steels
Very limited amount of particles in sodium
Low oxygen and hydrogen solubility
Very good wetting
IAEA
Sodium properties: Three main drawbacks
Very important reactivity with water
possible deleterious effects in Steam Generator Units
(SGU), in case of pipe rupture
Must be avoided or mitigated by design
• Selection of a modular SGU
Must be detected,
• Thanks to the production of hydrogen
• Risk of hydrogen explosion has to be mitigated
Important chemical reactivity with air
Can induce Na fire
Need inert zones and confinement
Need early detection
Opacity
Need specific equipments for under -sodium viewing and
measurements
IAEA
Lead-cooled Fast Reactor (LFR) outlook
• Limited experience (LBE-cooled
ALFA-class submarines in
Russian Federation)
• Resolve feasibility with respect
to components and
corrosion/erosion control
• Key technical issues:
Materials
Design features
High temperature applications: Operating
parameters (800 C)
IAEA
Gas-cooled Fast Reactor (GFR) outlook
• No experience
• Some benefit from VHTR
• Key technical issues:
SiC clad carbide fuel
Safety
Components and materials (850 C)
IAEA
Two main LMFR lay-outs
IAEA
EBR 1
1951: First SFR and first
electricity production from
nuclear
Historical records
IAEA
SFR – Experimental Reactors
Thermal
Power
(MW)
First
CriticalityShut down Country Comments
1946
1948
1950
1952
1954
1956
1958
1960
1962
1964
1966
1968
1970
1972
1974
1976
1978
1980
1982
1984
1986
1988
1990
1992
1994
1996
1998
2000
2002
2004
2006
2008
2010
2012
CEFR 65 2010 China
RAPSODIE 24 / 40 1967 1983 France
DFR (NaK) 75 1959 1977 GB
KNK 1 - KNK 2 60 1972 1991 Germany
FBTR 40 1985 India
PEC 120 Italy
JOYO 50 1977 Japan
BR 1 1955 Russia
BR 2 0,2 1956 1957 Russia
BR 5 - BR 10 05-oct 1958 2002 Russia
BOR 60 60 1968 Russia
CLEMENTINE (Hg) 0,02 1946 1964 USA
EBR 1 1,4 1951 1963 USA First FR electricity
LAMPRE 1 1961 1965 USA
EBR 2 60 1963 1993 USA
SEFOR 20 1969 1972 USA
FFTF 400 1980 1993 USA
1950 2000
IAEA
SFR - Demonstrators
Electrical
Power
(MW)
First
CriticalityShut down Country Comments
FERMI (EFBR) 100 1963 1972 USA
BN 350 150 (*) 1972 1999 Kazakhstan
PHENIX 250 1973 2009 France
PFR 250 1974 1994 GB
SNR 300 300 Germany Construction stopped 1992
MONJU 280 1994 Japan
CLINCH RIVER (CRBR) 350 USA Construction stopped
KALIMER 150 2028 Korea Project
ASTRID 600 2020 France Project
IAEA
SFR – Power Reactors
Electrical
Power
(MW)
First
CriticalityShut down Country Comments
BN 600 600 1980 Russia
SUPERPHENIX 1200 1985 1998 France
BN 800 800 2012 Russia
SPX 2 1500 France Project stopped
SNR 2 1400 Germany Project stopped
CDFR 1300 GB Project stopped
EFR 1500 EU Project stopped
PRISM 1245 USA Project stopped
PFBR 500 2013 India Construction
JSFR 1500 2025 ? Japan Project
KALIMER 600 600 2035 ? Korea Project
BN-S 1200 Russia Project
CFR 1000 1000 China Project
IAEA
SFR in operation: BN-600 in Russian Federation
In critical condition more than 213 000 hours;
Electricity production: about 120 billion kWh;
Average value of the load factor equal to 73.95%.
General parameters
Thermal power, MWth
Electric power, MWe
Design lifetime, year
1470
600
30
Primary circuit
Arrangement
Reactor vessel support
Primary and secondary coolant
Number of heat removal loops
Inlet/outlet core sodium temperature, С
Sodium flow rate, t/h
Pool-type
At the bottom
Sodium
3
377/550
25000
Core and fuel
Fuel
Max. fuel burnup, % h.a.
Diameter, mm
Height, mm
Uranium dioxide pellets
11.1
2058
1030
On 08.04.2011 BN-600 - the largest operating sodium-cooled fast
reactor in the world - celebrated the 31st anniversary since it was
connected to the grid
IAEA
SFR in operation (suspended): MONJU in Japan
Electricity Output : 280MWe (714MWt), Sodium Coolant, MOX Fuel Core
Primary sodium loop Water/steam system
Secondary sodium
Primarysodium
Intermediate Heat Exchanger(IHX)
Primary Circulating
pump
Core
Air cooler（AC)
Evaporator(EV)
TurbineGenerator
Feed water pump
Sea water pump
Condenser
Secondary sodium loop
Secondary Circulating
pump
Super heater(SH)
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
R/V ＴＢ
loop C
loop B
loop ATemperature, flowrate etc.
Primary sodium reactor vessel inlet/outlet: 529/397˚C, 5100 t/h/loopSecondary sodium IHX inlet/outlet: 325/505˚C, 3700 t/h/loopSteam at the turbine inlet: 483˚C, 12.5MPa, 1137 t/h
Electricity Output : 280MWe (714MWt), Sodium Coolant, MOX Fuel Core
Primary sodium loop Water/steam system
Secondary sodium
Primarysodium
Intermediate Heat Exchanger(IHX)
Primary Circulating
pump
Core
Air cooler（AC)
Evaporator(EV)
TurbineGenerator
Feed water pump
Sea water pump
Condenser
Secondary sodium loop
Secondary Circulating
pump
Super heater(SH)
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
R/V ＴＢ
loop C
loop B
loop A
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
SH
EVＩＨＸ
ACS
R/V ＴＢ
loop C
loop B
loop ATemperature, flowrate etc.
Primary sodium reactor vessel inlet/outlet: 529/397˚C, 5100 t/h/loopSecondary sodium IHX inlet/outlet: 325/505˚C, 3700 t/h/loopSteam at the turbine inlet: 483˚C, 12.5MPa, 1137 t/h
IAEA
Construction startOctober 1985
System start-up test (SST)ConstructionFunction test
Commercial operation
We are here.May 2011
SST beginningDecember 1992
Completion of equipment installation
May 1991
First connection to gridAugust 1995
Sodium leak accidentDecember 1995
First criticalityApril 1994
14 year 5 month SST suspension
Core confirmation testMay-July 2010
40% power confirmation test
Power rising test40%75%100%
Construction startOctober 1985
System start-up test (SST)ConstructionFunction test
Commercial operation
We are here.May 2011
SST beginningDecember 1992
Completion of equipment installation
May 1991
First connection to gridAugust 1995
Sodium leak accidentDecember 1995
First criticalityApril 1994
14 year 5 month SST suspension
Core confirmation testMay-July 2010
40% power confirmation test
Power rising test40%75%100%
SFR in operation (suspended): MONJU in Japan
IAEA
SFR in operation (suspended): JOYO in Japan
• Experimental fast reactor at Japan Atomic Energy
Agency’s O-arai Research and Development Center
• 100 MWth sodium cooled fast reactor.
• First criticality was achieved in April 1977
• As materials testing reactors, it has shown excellent
performance for more than 26 years.
• Role of the JOYO experimental
fast reactor:
Advancement of technology through
operation and experiment.
Conducting irradiation tests on fuel
and materials.
Validation of innovative technology
for development of future FBR.
IAEA
• 40 MWt (13.5 MWe) loop-type experimental fast reactor located in Kalpakkam.
• First criticality on 18 October 1985.
• Important works including PFBR shielding experiments, testing of transfer arm in air,
boron enrichment, post-irradiation examination of FBTR fuel after 125 GWd/t burnup,
structural integrity testing, and reprocessing of carbide fuel are being carried out.
SFR in operation: FBTR in India
• Construction, commissioning and
operation of FBTR have given
considerable amount of experience
and confidence
• FBTR will continue to be the
workhorse for the testing of metallic
fuels and advanced structural
materials being developed at IGCAR
for the next generation of fast
reactors.
IAEA
2002.8
2000.5
1998.10
2008.12
2009.8
Preparation of Site
Main building finished
Reactor block
installation finished
Commissioning of Phase A ended
SFR in operation: CEFR in China
Thermal power, MW 65
Electric power, net, MW 20
Reactor core,
Height, cm 45.0
Diameter equivalent, cm 60.0
Fuel (first loading, (Pu, U)O2, [UO2]
Pu, total, kg 150.3 239Pu, kg 97.7 235U- (enrichment), (first loading) kg (%) 42.6 (19.6%), [236.7 (64.4%)]
Linear power max, W/cm 430
Neutron flux, n/cm2·s 3.7×1015
IAEA
SFR in operation: BOR-60 in Russian Federation
• BOR-60 is used for:
Material tests;
Isotopes production (nickel-63, strontium-89, gadolinium-153 );
Tests of various equipments of fast reactors;
Heat and electricity production.
• In operation since more than 41 years.
• In December 2009, Rostechnadzor has given the license to the RIAR for further
operation of BOR-60 up to 31.12.2014.
Overall plan:
Thermal power, MW Up to 60
Electrical power, MW 12
Primary circuit:
Coolant Sodium
Coolant temperature, C
Core inlet
Core outlet
Up to 360
Up to 530
Coolant flowrate through reactor, m3/h Up to 1200
Reactor core:
Maximum neutron flux density, n·cm-2
·s-1
3.71015
Maximum core power density, kW/L 1100
Average neutron energy, MeV 0.45
Fuel UO2 or UO2-PuO2
Enrichment with 235
U, % 45-90
Maximum contents of Pu, % Up to 40
Enrichment with 239
Pu, % Up to 70
IAEA
SFR under Construction: 500 MWe PFBR in India
Full name: Prototype Fast Breed Reactor 500 MWe
Designer: BAHVINI, India
Reactor type: Fast Breeder Reactor
Coolant: Liquid sodium
Neutron Spectrum: Fast Neutrons
Thermal/Electrical Capacity: 1900 MWt/660 MWe
Fuel Cycle: closed fuel cycle
Salient Features: Passive decay heat
removal system; it does not require water
for emergency cooling in the case of an
accident; core catcher
Design status: 1 unit under construction in
Kalpakkam - India; commissioning in 2012-
2013
IAEA
SFR under Construction: BN-800 in Russian Federation
Bird’s-eye view of the reactor
compartment of the main building
Reactor thermal power, MW 2100
Unit electrical power, MW 880
Unit net efficiency, % 40
Operation life, year 40
Breeding ratio 1.04
IAEA
Fast Reactor Under Development
Large-size SFR
JSFR, Japan, 1500 MWe
BN-1200, Russian
Federation, 1200 MWe
KALIMER, Korea Republic of,
600 MWe
CFR, China, 1000 MWe
IAEA
DEMO and PROTO Fast Reactors under Development
ALLEGRO, France,
100 MWth
----Thermal
Insulation
12
95110Anchor safety
vessel
11
----Core04
5561Inner vessel05
----Transfer Arm06
----Large Rotatable
Plug
07
----SRP/Control Plug08
----IHX09
----Primary Pump10
3644.8Core support
structure
02
3476Grid Plate03
01
No.
116134Main vessel
CFBRPFBR
Weight in tComponent
10 09
07 08
06
04 05
03
02 01
11
12
13
72
5
Ø11950
CFBR, India, 500 MWe
ELSY, EU, 600 MWe
ASTRID, France,
600 MWe
IAEA
Small Modular Fast Reactors under Development
BREST, Russian
Federation, 300 MWe 4S, Japan, 10 MWe
SSTAR, USA,
10-100 MWe
PRISM, USA, 155 MWe SVBR, Russian
Federation, 100 MWe
HYPERION, USA, 25 MWe
TerraPower TWR, USA,
500 MWe
IAEA 39
• Name: European Lead-Cooled
System - ELSY
• Designer: Project developed in the
framework of EC
• Reactor type: Pool type
• Coolant: Pure Lead
• Plant Size: 600 MWe
• Design Goals and Characteristic
Features: MOX fuel, Spiral tube
steam generators, Removable
internals, Diversified DHR systems,
Thermal efficiency 40%
DEMO and PROTO Fast Reactors under Development
ELSY
IAEA
SM-FR under Development
BREST-OD-300
• Name: Bystriy Reactor Estestvennoy
Bezopasnosti (Fast Reactor Natural Safety) -
BREST
• Designer: RDIPE, Russian Federation
• Reactor type: Liquid Metal Cooled Reactor
• Coolant: Lead
• Plant Size (Thermal capacity): 700 MWt
• Plant Size (Electrical Capacity): 300 MWe
• Design Goals and Characteristic Features:
PuN-UN Fuel Material, Fuel Cycle 10 Months,
High level of inherent safety due to natural
properties of fuel, core and cooling design.
40
IAEA
SM-FR under Development
SVBR-100
• Name: Svinetc-Vismuth Bystriy Reactor
(Lead-Bismuth Fast Reactor) / SVBR-100
• Designer: AKME, Russian Federation
• Reactor type: Liquid Metal Cooled Reactor
• Coolant: Lead-Bismuth
• Plant Size (thermal Capacity) : 280 MWt
• Plant Size (Elctrical Capacity): 106 MWe
• Design Goals and Salient Features: UO2
Fuel Material, Fuel Enrichment 16.5%, Fuel
Cycle 8 years, Closed Fuel Cycle with MOX,
operation in a self-sufficient mode
41
IAEA
SM-FR under Development
SSTAR (ANL, USA)
42
CLOSURE HEAD
CO2 INLET NOZZLE
(1 OF 4)
CO2 OUTLET NOZZLE
(1 OF 8)
Pb-TO-CO2 HEAT
EXCHANGER (1 OF 4)
ACTIVE CORE AND
FISSION GAS PLENUM
RADIAL REFLECTOR
FLOW DISTRIBUTOR
HEAD
FLOW SHROUDGUARD
VESSEL
REACTOR
VESSEL
CONTROL
ROD
DRIVES
CONTROL
ROD GUIDE
TUBES AND
DRIVELINES
THERMAL
BAFFLE
• Name: small, sealed, transportable,
autonomous reactor – SSTAR
• Designer: ANL, USA
• Reactor type: Pool type
• Coolant: Lead
• Plant Size: 19.8 MWe
• Design Goals and Characteristic
Features: Transuranic nitride
(TRUN) Fuel, Core Life Time 30
years, Supercritical CO2 Brayton
Cycle, Net Plant Efficiency 44%
IAEA
SM-FR under Development
Hyperion
• Name: Hyperion Power Module
• Designer: Hyperion Power
Generation, Inc., USA
• Reactor type: Liquid Metal Cooled
Reactor
• Coolant: Liquid Metal (Pb-Bi)
• Plant Size: 70 MWt / 25 MWe
• Design Goals and Salient Features:
Uranium Nitride Fuel Material,
Enrichment 19.75%, Fuel Cycle 10
years, Transportable Factory Fueled
Design
43
IAEA
MBIR (Russia)
• Na-cooled Research Fast
Reactor aimed at in-pile tests
of new types of fuel,
structural materials and
various FR coolants (Na, Pb,
Pb-Bi, etc.)
• Start-up of MBIR is
scheduled in 2019.
44
Lay-out of the MBIR reactor vessel
and its experimental channels
IAEA
BASIC CHARACTERISTICS OF THE MBIR REACTOR
45
Parameter Value
Thermal power, MW ~150
Electric power, MW ~40
Maximum neutron flux density, n·cm-2·s-1 ~6.0·1015
Driven fuel Vi-pack-MOX, (PuN+UN)
Test fuel Innovative fuels, MA fuels and targets
Core height, mm 600
Maximum core power density, kW/l 1100
Maximum neutron fluence per year, n·cm-2 ~ 1·1023 (up to 45 dpa)
Design lifetime, year 50
Number of autonomous test loops with different
coolants up to 4
Total number of experimental subassemblies and
target devices for radioisotope production
up to 12 (core)
up to 5 (radial shielding)
Number of experimental channels up to 3 (core)
Number of experimental horizontal channels up to 6 (outside reactor vessel)
Number of experimental vertical channels up to 8 (outside reactor vessel)
IAEA
MYRRHA by SCK.CEN (Belgium)
Fast
Neutron
Source
Spallation Source
Lead-Bismuth
coolant
Multipurpose
Flexible
Irradiation
Facility
IAEA
International Initiatives
Generation IV International Forum - GIF 4th Generation Nuclear Systems for
sustainable energy development
E.U.
China
Russia
Technical maturity around 2030
Steady progress
• Economic Competitiveness
• Safety and reliability
Significant progress :
• Waste minimisation
• Resource saving
• Security : non proliferation, physical protection
Opening to other applications :
• High temperature heat for industry
• Hydrogen, drinking water
IAEA
International Initiatives
IAEA - INPRO
The International Project on
Innovative Nuclear Reactors
and Fuel Cycles (INPRO) was
established in 2000
To help ensure that nuclear energy
is available to contribute to meeting
the energy needs of the 21st
century in a sustainable manner
To bring together technology
holders, technology users and other
stakeholders to consider jointly the
national and international actions
required to achieve desired
innovations in nuclear reactors and
fuel cycles.
IAEA
ESNII - The European Sustainable Nuclear
Industrial Initiative
SFR Prototype Astrid
250-600 MWe
2008 2012 2020
LFR
SFR
GFR
Supporting infrastructures, research facilities, irradiation facilities
& fuel manufacturing and reprocessing facilities
LFR demonstrators
MYRRHA
ALFRED
Reference (proven) technology
Alternative technology Allegro GFR Demo •Test bed of GFR
technologies
•Innovative fuel
•MA transmutation
•Coupling to heat applications
MOX fuel fab unit
MA fuel micropilot
IAEA 50
Key Elements for Future Development of Fast
Reactors Technology
IAEA 51
Key Elements for FR Development
Driving force and peculiarity is sustainability: Natural resources preservation
Waste minimization
but maintaining/improving
Safety (also in light of Fukushima accident)
Economical competitiveness
Proliferation Resistance
Identify technological gaps to achieve the objectives in those
areas represents a key step for future developments
IAEA 52
Innovation to ensure that enhanced requirements for
performance, safety and costs are met research
and technology development
Clear path from research and technology
development to deployment
Limited resources (economic/financial crisis +
industrial and governmental focus on existing and
near-term deployment reactors): need to optimize
human and financial resources collaboration
Key Elements for FR Development
IAEA
Advanced Simulation & Modelling (e.g. multi-physics and multi-
scale computer codes) with higher level of precision, key for:
design optimization (e.g. reduce nominal pick temperatures);
drastically reducing uncertainty margins;
narrowing down the needs of expensive experimental tests (mock-ups,
T/H and safety experiments, etc.)
Data and computer code verification, validation, and
qualification (V&V&Q) through theoretical and experimental
benchmarks, including severe accident analyses
Comparative assessments of feasibility, performance, and
safety characteristics
Technological choices / concept selection
53
Technical Areas for R&D and Innovation
IAEA 54
Technical Areas for R&D and Innovation
Advanced structural materials
Innovative fuels (including MA-bearing fuels)
Core performance
Primary and secondary system simplification
Compact Heat Exchangers
New Power Conversion Systems
Coolant technologies and Advanced instrumentation
In-service Inspection
Safety (reactivity effects, passive systems, seismic behaviour, etc.)
& Security
IAEA 55
General Framework of
Safety of Fast Reactors
IAEA 56
The Accident which occurred at the
Fukushima Dai-ichi NPP on March
2011 has led a renewed global
attention and concern on the safety of
current and future nuclear energy
systems
Significant efforts are nowadays
devoted to identify the main lessons
learned from this event
On the basis of the lessons learned, a
review of safety design approaches
and features of any nuclear system is
considered necessary in view of future
developments of nuclear power
The Fukushima Accident
IAEA 57
Even current and future fast neutron reactors have to take into account the
main lessons learned from the Fukushima accident
In particular, extreme external events which may potentially lead to severe
accidental scenarios such as Station Black Out, Loss of Ultimate Heat Sink as
well as others have to be considered and analysed
The introduction of advanced technical solutions and provisions for the
prevention and mitigation of these scenarios (beyond design basis accidents)
is a key factor to further increase the safety level
Even if it is possible to identify common approaches, the safety characteristics
of fast reactors are rather different than that of thermal reactors, and they
require therefore appropriate and specific solutions
Implications on Fast Reactors
IAEA 58
Main Safety Characteristics of LMFRs
LMFR have a number of favourable safety characteristics with respect to other
nuclear systems (in particular L/HWR), i.e.:
Easy to operate:
No pressure at the primary circuit,
High thermal inertia,
Control by rod position (no Xe effect, no need of soluble neutron poison).
Radioprotection level higher than in LWR;
Few effluents;
High thermal and electrical conductivity;
High thermal efficiency;
Large coolant boiling margin;
Natural convection.
IAEA 59
Main Safety Characteristics of LMFRs
However, LMFR have also characteristics representing design
challenges for safe operation, i.e.:
High power density: need to provide adequate heat removal under all
circumstances;
Power variation due to neutron leakage at the core boundaries: core
reactivity is very sensitive to core geometry;
Core coolant void worth is typically positive;
Core is not in the most reactivity configuration, with a core inventory
of many critical masses: fuel relocation might significantly increase
reactivity, potentially leading to very high power generation (1000s x
nominal) Core Disruptive Accident
IAEA 60
Main Safety Characteristics of LMFRs
FRs intrinsic safety-related features are very different than that of thermal
reactors:
Intrinsic characteristics of the fast spectrum (shorter life time of prompt
neutron, small number of delayed neutrons, high power density)
Different coolants and plant configurations
Main concerns are related to accident scenarios (e.g. UTOP, ULOF,
USTOP, etc.) which can potentially bring to core disruption events
Solutions based on both intrinsic physical mechanisms (negative reactivity
feedbacks) and engineered safety systems have to be provided in order to
assure the required level of safety.
IAEA 61
Defence-in-Depth: multiple redundant active and passive safety systems
Two (in GENIV also three…) redundant and independent
shutdown systems: diverse, robust and reliable
Multiple coolant pumps
Redundancy and diversification for DHRS
Multiple barriers to the release of radioactive materials:
Cladding on fuel pins
Primary coolant system boundary
Containment building
Negative power and temperature reactivity feedback coeff.
IAEA 62
Inherent safety characteristics: preventing severe consequences from unprotected accidents
Based on fundamental phenomena such as thermal expansion,
buoyancy-driven flow, and gravity (reliability to be quantified)
Mainly address ULOF, ULOHS, and inadvertent withdrawal of
CR(s) resulting in a transient overpower accident (UTOP)
The focus is on the three main conditions for safe operation of
the reactor:
Avoid large uncontrolled increases in core power, by means of favorable
reactivity feedbacks;
Avoid insufficient cooling of the reactor core, by means of natural
circulation cooling;
Avoid rearrangement of fuel that would lead to energetic events, (solid or
molten core compaction) by core or Sub-Assembly design.
IAEA 63
Beyond ULOF, ULOHS, UTOP
E.g. UTOP where all control rods are uncontrollably withdrawn
from the core
Even inherent safety features are unable to prevent temperature
increases, coolant boiling, fuel melting and fuel pin failure
Probability of occurrence less than 10-6 per reactor.year
Mitigation:
preservation of the mechanical integrity of the reactor vessel
favorable dispersal of the molten fuel to prevent energetic recriticalities
and to maintain core coolability.
IAEA 64
Japanese government report to the IAEA - 28 key points
grouped in the following 5 Groups:
1. Strengthen preventive measures against a severe
accident
2. Enhancement of measures against severe accidents
3. Enhancement of nuclear emergency response
4. Reinforcement of safety infrastructure
5. Raise awareness of safety culture
Lessons learned from Fukushima accident (1/3)
IAEA 65
Lessons learned from Fukushima accident (2/3)
Strengthen measures against earthquakes and tsunamis and, in general, extreme
external events (with reconsideration of their magnitude during plant design)
Combination of events: in particular combination of hazards and combinations of
external hazards with accidents
Identification of cliff-edge effects associated to hazards
Emergency power supply: diversity to the extent practicable and redundancy for
suppressing common cause failure including external events
Decay heat removal system:
reactor cooling even under loss of all AC power supply
utilization of passive heat removal capability for DEC
diversity to the extent practicable and redundancy for suppressing common cause failure
including external events
Ultimate heat sink: diversity of the ultimate heat sinks for the decay heat transfer
IAEA 66
Lessons learned from Fukushima accident (3/3)
Improvement of the containment
Reduction of potential bypass
Independence of confinement barriers
Environmental impact of chemically hazardous material (e.g., sodium)
Fuel storage systems:
adequate heat removal
status monitoring even under DEC including the loss of all AC power supplies
External hazards:
due consideration of loss of all AC power supplies following the extreme external hazards
seismic events may be accompanied by subsequent events
Means of radiation monitoring: adequate radiation monitoring in DEC
Design extension conditions: designs/provisions for Prevention and Mitigation of the
severe accident consequences.
IAEA
IAEA Action Plan on Nuclear Safety
11 March 2011: Great East Japan Earthquake
June 2011: IAEA Ministerial Conference on Nuclear
Safety
Declared a number of agreed points that should direct the
process of learning and acting upon lessons to strengthen
nuclear safety, emergency preparedness and radiation
protection
Requested that IAEA prepares an Action Plan building upon
the declarations and recommendations of working groups
67
IAEA
Development of IAEA Action Plan
June to Sept. 2011 – Draft action plan developed
• Produced in conjunction with Member States
• Considered the points of
ministerial report,
working group reports,
IAEA fact finding mission to Fukushima, and
views of the International Nuclear Safety Group (INSAG)
September 2011 – Action plan approved by IAEA Board
of Governors
68
IAEA
IAEA Action Plan on Nuclear Safety
12 Actions:
Safety Vulnerabilities
Peer Reviews
EPR
Regulatory Bodies
Operating Organisations
IAEA Safety Standards
Legal Framework
Embarking countries
Capacity Building
Protection of People & Environment
Communication, Research & Development
IAEA
IAEA Action Plan: progress so far
Nuclear safety action team formed
Implementation strategy under development
Status website established:
http://www.iaea.org/newscenter/focus/actionplan
Vulnerability assessment method created
Safety Standards action plan created
Draft review of document identifying gaps in IAEA safety standards with
respect to Fukushima lessons learned so far, available at
http://www-ns.iaea.org/committees/comments/default.asp?fd=1114
70
IAEA
Action Plan Next Steps
March 2012: International experts’ meeting at IAEA-
HQ to discuss Fukushima lessons
June 2012: Report to Board of Governors
September 2012: Progress on the implementation of
the Action Plan will be reported to the Board of
Governors
Next General Conference 2012
IAEA lessons learned report by the end of 2012
71
IAEA 72
IAEA Activities in support of
Fast Reactors Development
IAEA
Main activities of the IAEA Programme on Fast Reactor (1/2)
Organize regular Topical Technical Meetings for in-depth information exchange related to development, design, construction and operation of nuclear power plants with Fast Reactors (FR), as well as to R&D on Accelerator Driven Systems (ADS)
Organize Large Conferences on different aspects of FR and ADS RTD (e.g. “Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities” – FR09, Kyoto December 2009; FR013 – Paris, spring 2013)
Establish a forum for broad exchanges on technical requirements for and characteristics of 4th Generation Fast Reactor Systems, also in collaboration with INPRO
73
IAEA
Main activities of the IAEA Programme on Fast Reactor (2/2)
Carry out Coordinate Research Projects (CRPs) of common
interest to the TWG-FR Member States in the field of FRs and
ADS
Secure Training and Education in the field of fast neutron
system physics, technology and applications
Provide support to IAEA Nuclear Safety and Security
Department for preparation of fast reactor Safety standards /
requirements / guides
Task #1 of the project “Support for Fast Reactor RT&D&D”:
Support Fast Reactor data retrieval and knowledge
preservation activities in MSs
74
IAEA
The IAEA Technical Working Group on
Fast Reactors
75
Members of the IAEA Technical Working Group on Fast Reactors
Full Members
Belarus Brazil
China France
Germany India
Italy Japan
Kazakhstan Korea, republic of
Netherlands Russian Federation
Switzerland Ukraine
UK USA
OECD/NEA European Commission
Observers
Argentina Belgium
Spain Sweden
Participants in the
44th Annual Meeting of the TWG-FR,
Institute of Atomic Energy (CIAE), Beijing,
China, 23-27 May 2011
Full Members Observers
Members of the IAEA Technical Working Group on Fast Reactors
IAEA 76
CRPs Recently Completed
IAEA 77
CRP on “Analytical and Experimental Benchmark
Analyses of Accelerator Driven Systems (ADS)”
Participants
Argentina Hungary
Belarus Italy
Brazil Japan
China Netherlands
Poland Pakistan
Russian Federation Spain
Belgium Sweden
France Ukraine
Germany USA
Advance efforts towards designing a
demonstration facility by providing
information exchange and collaborative
research framework
Improve the present understanding of the
coupling of ADS spallation sources with
multiplicative sub-critical nuclear system
Fast
Neutron
Source
Spallation Source
Lead-Bismuth
coolant
Multipurpose
Flexible
Irradiation
Facility
CRP on Analytical and Experimental Benchmark Analyses of
Accelerator Driven Systems (ADS)
IAEA
CRP on Analyses of, and Lessons Learned from the Operational
Experience with Fast Reactor Equipment and Systems
• Preserve the feedback from commissioning, operation, and decommissioning
experience of experimental and power sodium cooled fast reactors
• Retrieve, assess, review and archive of all the relevant documentation and
information
• Enable easy access to the information from this feedback
• Produce lessons-learned, synthesis reports of lessons learned and
recommendations from the commissioning, operation, and decommissioning
of experimental and power sodium cooled fast reactors
“Analyses of and lessons learned from the operational
experience with fast reactor equipment and systems”
Participants
France India
Japan Korea, Republic of
Russian Federation
IAEA
“Control Rod Withdrawal and Sodium
Natural Circulation Tests Performed
during the PHENIX End-of-Life
Experiments”
Participants
China France
India Japan
Korea, Republic of Russian Federation
Switzerland USA
CRP on Control Rod Withdrawal and Sodium Natural Circulation Tests
Performed During the PHENIX End-of-Life Tests (special session at ICAPP-12)
Experimental benchmark exercises (preparatory
analyses, blind calculations, and post-
experiment analyses) based on the data
obtained during the PHENIX End-of-Life tests
V&V of methods and codes currently employed
in the field of FR neutronics, thermal hydraulics
and plant dynamics to achieve enhanced safety
IAEA
CP on Integrated Approach for the Modelling of Safety Grade
Decay Heat Removal System for LMR (Report under preparation)
Reference Design: 500 MWe pool type PFBR
Detailed analysis of a DHR system using
different codes and modelling approaches to
inter-compare the results obtained (7 case
studies for different conditions)
INPRO Collaborative Project:
“Integrated Approach for the Modelling of Safety Grade Decay Heat
Removal System for Liquid Metal Reactors"”
Participants
China EU/JRC
India Korea, Republic of
Russian Federation
IAEA 81
CRPs to be Completed and to be Launched
IAEA
“Benchmark Analyses of Sodium Natural
Convection in the Upper Plenum of the MONJU
Reactor Vessel”
Participants
China France
India Japan
Korea, Republic of Russian Federation
USA
CRP on Benchmark Analyses of Sodium Natural Convection in the Upper
Plenum of the MONJU Reactor Vessel (special session at NURETH-14)
Validation of CFD methods and turbulence models
based on Na thermal stratification measurements
performed in MONJU during a reactor turbine trip
test conducted in December 1995 in the course of
the original start-up experiments
Thorough assessment of the calculation versus
measured data comparisons
IAEA
“Benchmark Analyses of an EBR-II Shutdown Heat Removal
Test”
Expression of Interest
China Germany
Italy India
Japan Korea, republic of
Netherlands Russian Federation
Sweden Switzerland
USA
CRP on Benchmark Analysis of an EBR-II
Shutdown Heat Removal Test (to be launched in 2012)
A comprehensive testing program (45 tests !) conducted
between 1984 and 1987 A unique set of whole-plant
safety tests that demonstrated the potential for SFR to
survive severe accident initiators with no damage
Two EBR-II loss of flow tests chosen for this IAEA CRP:
SHRT-17, the most severe of the loss of flow with
scram tests
SHRT-45, the most severe of the loss of flow without
scram tests
IAEA 84
New CRP on SFR: Sodium Properties, Sodium Facility Design and
Safety Guidelines (to be launched in 2012-2013)
This CRP is proposed by France and it is
intended to address the needs of
standardization of Na physical and
chemical properties, the main rules for
designing experimental facilities, good
practices and safety guidelines
The CRP – making available validated
data and correlations for Na coolant - will
also improve the modelling and simulation
capabilities in various fields of SFR
technology
The outputs of this CRP will contribute to
an improvement of the future benchmark
exercises and of the design of sodium
facilities and their safe operation.
IAEA
CRP on Source Term for Radioactivity Release under FR Core
Disruptive Accident (CDA) Conditions (to be launched in 2012)
Demonstrate through numerical simulations of
FPs transport mechanisms that in future FBRs
the radioactivity release to the environment is
very low even in the extreme case of CDA
Under whole core accident, the fission products
and radioactive sodium are the basic source for
the radioactivity release
Reference design for the
safety analysis:
500 MWe pool type PFBR
??
?
? ?
?
IAEA 86
Conferences, Workshops & Technical Meetings
IAEA 87
FR09 – International Conference on Fast
Reactors and Related Fuel Cycles:
Challenges and Opportunities, Kyoto, 7-
11 December 2009
Major Conferences
AccApp'11 - International Conference on Nuclear
Research Applications and Utilization of
Accelerators, Oak Ridge, Tennessee, USA, 3-7
April 2011
IAEA 88
FR13 – International Conference on
Fast Reactors and Related Fuel
Cycles - Paris, 3 – 7 March 2013 (co-
organized with IAEA Nuclear Fuel
Cycle Section and CEA/SFEN-
France)
AccApp’13 - International Conference
on Nuclear Research Applications
and Utilization of Accelerators, Ghent,
USA, 26-27 July 2013
Major Conferences
IAEA 89
TM on “Fast Reactor Physics and Technologies”, Kalpakkam,
14-18 November 2011
GIF-IAEA/INPRO Workshop on Safety Aspects of Sodium
Fast Reactors, Vienna, 30 Nov. – 1 Dec. 2011
TM (in cooperation with NKM Unit) on the “IAEA’s Fast
Reactor Data Retrieval and Knowledge Preservation
Initiative”, Vienna, 6-8 December 2011
TM on “Fast Reactors In-service Inspection and Repair: Status
and Innovative Solutions”, Vienna, 19-20 December 2011
TM on “Innovative Heat Exchanger and Steam Generator
Designs for Fast Reactors”, Vienna, 21-22 December 2011
Technical Meetings & Workshops
IAEA 90
Technical Meetings & Workshops
TM on “Innovative Fast Reactor Designs with Enhanced
Negative Reactivity Feedback Effects”, Vienna, 27-29 February
2012
TM to “Identify Innovative Fast Neutron Systems Development
Gaps”, Vienna, 29 February – 2 March 2012
GIF-INPRO Interface Meeting, Vienna, 6 – 7 March 2012
TM on “Impact of Fukushima event on current and future FR
designs”, Dresden, 19 - 23 March 2012
IAEA-JAEA Workshop on Safety of SFR, Tsuruga, Japan, 11 –
13 June 2012
IAEA 91
Technical Meetings & Workshops
Fourth Research Coordination Meeting of the CRP on
"Benchmark analyses of sodium natural convection in the upper
plenum of the MONJU reactor vessel”, Tsuruga, Japan, 16 – 20
April 2012
First Research Coordination Meeting of the CRP on "Analyses
of Fast Reactor Safety Tests Conducted in EBR-II”, Argonne,
USA, 18 – 19 June, 2012
45th TWG-FR Annual TM, Argonne, 20 – 22 June 2012
TM on Construction & Commissioning of SFR, Kalpakkam,
India, November 2012
IAEA 92
Main Deliverables
IAEA
Forthcoming TWG-FR Technical Publications (1/2)
Status of Fast Reactor Research and Technology
Development (850 pages IAEA TECDOC ! in print): Background and overview
Operating experience with SFR
Sodium-cooled FR Designs
HLM-cooled FR Designs
Gas-cooled FR Designs
Status of FR core R&D
Reactor plant engineering technology development
Reactor safety design and analysis
National strategies, international initiatives, public acceptance and final
remarks
93
IAEA
Technical Reports closed to Publication
Liquid metal coolants for Fast
Reactors: reactors cooled by sodium,
lead and lead-bismuth eutectic (in
print)
Design Features and Operating
Experiences of Experimental Fast
Reactors (in print)
Proceedings of FR09, Kyoto,
December 2009 (in print)
94
IAEA
Technical Reports and NES in Preparation
BN-600 Hybrid Core Benchmark Analysis: methods to reduce
calculation uncertainties of the LMFR reactivity effects (under
final editing)
Benchmark analyses on the Natural Circulation Test
Performed During the PHENIX End-of-Life Experiments (under
final editing)
Status Report of Accelerator Driven Systems for waste
transmutation and energy production (under final editing)
Special issue of Nuclear Engineering & Design Journal
devoted to the outcomes of the IAEA TM on Physics and
Technology of Fast Reactors (papers available and under
review)
95
IAEA
Technical Reports and NES in Preparation
Final Report of the CRP on Analytical and Experimental
Benchmark Analyses of Accelerator Driven System
Final Report of the CRP on Lessons Learned from the
Operational Experience on Fast Reactors (editing just started)
Final Reports of the CRP on Control Rod Withdrawal and
Sodium Natural Circulation Tests Performed During the
PHENIX End-of-Life Tests (first report under review)
Final Report of the CRP on Benchmark Analyses of Sodium
Natural Convection in the Upper Plenum of the MONJU
Reactor Vessel
96
IAEA 97
FR Project WEB-site:
http://www.iaea.org/NuclearPower/FR/
IAEA 98
TWG-FR WEB-site
http://www.iaea.org/NuclearPower/Technology/TWG/TWG-FR/
IAEA
Efforts to preserve
knowledge and past
experience gathered in the
operation of fast reactors
and FR and ADS R&D
programmes
IAEA Fast Reactor and
ADS Data Bases jointly
managed by NPTDS/FR
Project and Nuclear
Knowledge Management
(NKM) Unit
Fast Reactor Knowledge Preservation: FR Data Base
http://www.iaea.org/inisnkm/nkm/aws/frdb/
IAEA 100
NKM Unit developed a Fast
Reactor Knowledge
Organization System (FR-
KOS): IT system to retrieve
information stored in an
international data base
The NPTDS/FR Project
collaborates with the NKM
Unit to update the system
and collect data and info to
be uploaded into the system
First version released to MSs
for testing and stimulating
contributions to FR-KOS
FRs Knowledge Preservation: FR-KOS System
https://nkm.iaea.org/nkm1
IAEA
Conclusions
Nuclear power has the potential to play a prominent role for the future generations energy supply
needs.
Almost all nuclear reactors under operation are thermal reactors, which don’t allow a complete
utilization of natural resources, posing concerns on the future availability of resources
Fast reactors enhance sustainability of nuclear energy utilization. The characteristics of the fast
spectrum and the closed fuel cycle guarantee a potential energy supply for thousand years and a
more suitable waste management
Fast reactors technology has been brought to a high level of technical maturity in the last decades
by the design and construction of experimental and prototype reactors, and several construction
projects are currently on going
Several countries are engaged in the development of innovative FR concepts, and therefore
important research efforts are worldwide devoted to cover technology gaps and improve safety
features, especially in the light of the Fukushima event.
In order to promote cooperation, international initiatives have been established in last years (GIF,
INPRO, ESNII)
The IAEA supports Member States Activities by providing a forum for information exchange and
implementing collaborative research programmes. The Agency’s activities in this field are carried
out in the framework of the TWG-FR
101
IAEA 102
http://www.iaea.org/NuclearPower/FR/
…Atoms for Peace
Thanks for Your Attention !
IAEA 103
Back
IAEA
CFR-1000 diagram
• Name: CFR-1000
• Designer: CIAE, China
• Reactor type: Pool type
• Coolant: Sodium
• Plant Size: 2500 MWt - 1000
MWe
• Design Goals and Salient
Features: MOX fuel, Breeding
Ratio targeted at 1.2, Plant
Design Life of 40y, Na-Na-H2O
loops with 3 circuits of primary
and secondary loop
Large Size Fast Reactor Under Development
CFR-1000
IAEA
• Name: BN-1200
• Designer: OKBM, Russian Federation
• Reactor type: Pool type
• Coolant: Sodium
• Plant Size: 2900 MWt – 1220 MWe
• Design Goals and Salient
Features: MOX Fuel type, Breeding
ratio targeted at 1.2, Gross thermal
efficiency of 42%, 4 heat removal
loops, Plant Design Life of 60 years
Large Size Fast Reactor Under Development
BN-1200
IAEA
MBIR (Russia)
• Na-cooled Research Fast
Reactor aimed at in-pile tests
of new types of fuel,
structural materials and
various FR coolants (Na, Pb,
Pb-Bi, etc.)
• Start-up of MBIR is
scheduled in 2019.
106
Lay-out of the MBIR reactor vessel
and its experimental channels
IAEA
BASIC CHARACTERISTICS OF THE MBIR REACTOR
107
Parameter Value
Thermal power, MW ~150
Electric power, MW ~40
Maximum neutron flux density, n·cm-2·s-1 ~6.0·1015
Driven fuel Vi-pack-MOX, (PuN+UN)
Test fuel Innovative fuels, MA fuels and targets
Core height, mm 600
Maximum core power density, kW/l 1100
Maximum neutron fluence per year, n·cm-2 ~ 1·1023 (up to 45 dpa)
Design lifetime, year 50
Number of autonomous test loops with different
coolants up to 4
Total number of experimental subassemblies and
target devices for radioisotope production
up to 12 (core)
up to 5 (radial shielding)
Number of experimental channels up to 3 (core)
Number of experimental horizontal channels up to 6 (outside reactor vessel)
Number of experimental vertical channels up to 8 (outside reactor vessel)
IAEA
MYRRHA by SCK.CEN (Belgium)
Accelerator
(600 MeV – ≤ 4 mA proton)
Reactor
•subcritical mode (~85 MWth)
•Critical mode (~100 MWth)
