sodium fast reactors systems and components (part 2) · pdf file ·...
TRANSCRIPT
Dr Christian LATGE
Nuclear Technology Department
Nuclear Energy Division
CEA Cadarache
13108 Saint Paul lez Durance France
Phone +33 4 42 25 44 71
Fax +33 4 42 25 78 78
e-mail christianlatgeceafr
IAEA Education ampTraining Seminar
on
Fast Reactor Science and Technology
CNEA Bariloche Argentina October 1 ndash 5 2012
Sodium Fast Reactors
Systems and components (Part 2)
Mechanical pumps
Pump with vertical shaft impeller hydrostatic
bearing hang to the reactor slab (with lateral
inlet and axial outlet)
P1 pressure at pump outlet (highest value in the primary
loop)
P2 pressure at core outlet (P2=P1-DPcore)
P3 pressure in the cold vessel (P2=P2-DPexchanger H is
representative of this DP)
All these pressure drops depends of flowrates
Needs for design
Pump supporting devices
Sizing of impeller
Evaluation of cavitation risk
Shatf Guiding device
Tightness
For SPX Qualification in water (scale 1)
Pressure distribution in the primary vessel
Cavitation
Cavitation is the spontaneous production of vapour bubbles
in the liquid phase due to the fact that local pressure
becomes lower than saturation vapour pressure
Every flow is controlled by the Bernoulli equation
P + gH+ 12 ρV2-ΔP = cte
P static pressure
H height denfoncement
V flow velocity
ρ masse volumique
ΔP pressure drop
By simplification if we assume H =cte and neglecting ΔP
then P + 12ρV2=cte
If velocity increases pressure decreases and can reach the
saturation vapour pressure production of vapour bubles
and consequences (lower efficiencymechanical impact
erosion noise)
Properties of sodium
Electrical resistivity in the liquid state
e ( m) = 61405 10-8 + 35047 10-10 + 56885 10-14 2 + 166797 10-17 3
Consequences
The conductive properties of sodium are used in instrumentation flow rate measurements electromagnetic pumps Na leak detection etc
Electromagnetic pumps basic principle
Conduction pump current is introduced by electrodes and
conducting coolant (ie Na) circulates in the pipe thanks to
magnetic field
Induction pump current is generated directly in the liquid by
induction (a variable electromagnetic field is generated in time
and space in the liquid)
Laplace equation
Conduction pump
Induction pump
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Mechanical pumps
Pump with vertical shaft impeller hydrostatic
bearing hang to the reactor slab (with lateral
inlet and axial outlet)
P1 pressure at pump outlet (highest value in the primary
loop)
P2 pressure at core outlet (P2=P1-DPcore)
P3 pressure in the cold vessel (P2=P2-DPexchanger H is
representative of this DP)
All these pressure drops depends of flowrates
Needs for design
Pump supporting devices
Sizing of impeller
Evaluation of cavitation risk
Shatf Guiding device
Tightness
For SPX Qualification in water (scale 1)
Pressure distribution in the primary vessel
Cavitation
Cavitation is the spontaneous production of vapour bubbles
in the liquid phase due to the fact that local pressure
becomes lower than saturation vapour pressure
Every flow is controlled by the Bernoulli equation
P + gH+ 12 ρV2-ΔP = cte
P static pressure
H height denfoncement
V flow velocity
ρ masse volumique
ΔP pressure drop
By simplification if we assume H =cte and neglecting ΔP
then P + 12ρV2=cte
If velocity increases pressure decreases and can reach the
saturation vapour pressure production of vapour bubles
and consequences (lower efficiencymechanical impact
erosion noise)
Properties of sodium
Electrical resistivity in the liquid state
e ( m) = 61405 10-8 + 35047 10-10 + 56885 10-14 2 + 166797 10-17 3
Consequences
The conductive properties of sodium are used in instrumentation flow rate measurements electromagnetic pumps Na leak detection etc
Electromagnetic pumps basic principle
Conduction pump current is introduced by electrodes and
conducting coolant (ie Na) circulates in the pipe thanks to
magnetic field
Induction pump current is generated directly in the liquid by
induction (a variable electromagnetic field is generated in time
and space in the liquid)
Laplace equation
Conduction pump
Induction pump
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Cavitation
Cavitation is the spontaneous production of vapour bubbles
in the liquid phase due to the fact that local pressure
becomes lower than saturation vapour pressure
Every flow is controlled by the Bernoulli equation
P + gH+ 12 ρV2-ΔP = cte
P static pressure
H height denfoncement
V flow velocity
ρ masse volumique
ΔP pressure drop
By simplification if we assume H =cte and neglecting ΔP
then P + 12ρV2=cte
If velocity increases pressure decreases and can reach the
saturation vapour pressure production of vapour bubles
and consequences (lower efficiencymechanical impact
erosion noise)
Properties of sodium
Electrical resistivity in the liquid state
e ( m) = 61405 10-8 + 35047 10-10 + 56885 10-14 2 + 166797 10-17 3
Consequences
The conductive properties of sodium are used in instrumentation flow rate measurements electromagnetic pumps Na leak detection etc
Electromagnetic pumps basic principle
Conduction pump current is introduced by electrodes and
conducting coolant (ie Na) circulates in the pipe thanks to
magnetic field
Induction pump current is generated directly in the liquid by
induction (a variable electromagnetic field is generated in time
and space in the liquid)
Laplace equation
Conduction pump
Induction pump
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Properties of sodium
Electrical resistivity in the liquid state
e ( m) = 61405 10-8 + 35047 10-10 + 56885 10-14 2 + 166797 10-17 3
Consequences
The conductive properties of sodium are used in instrumentation flow rate measurements electromagnetic pumps Na leak detection etc
Electromagnetic pumps basic principle
Conduction pump current is introduced by electrodes and
conducting coolant (ie Na) circulates in the pipe thanks to
magnetic field
Induction pump current is generated directly in the liquid by
induction (a variable electromagnetic field is generated in time
and space in the liquid)
Laplace equation
Conduction pump
Induction pump
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Electromagnetic pumps basic principle
Conduction pump current is introduced by electrodes and
conducting coolant (ie Na) circulates in the pipe thanks to
magnetic field
Induction pump current is generated directly in the liquid by
induction (a variable electromagnetic field is generated in time
and space in the liquid)
Laplace equation
Conduction pump
Induction pump
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Interest in Annular Linear Induction Pump for ASTRID
ALIP in ASTRIDrsquos intermediate circuit
Primary circuit circuit Power Conversion System
either Rankine Steam Cycle
or Brayton Gas Cycle
primary circuit intermediate circuit
pump
generator
core
network
Connection to the
primary vessel (IHX)
4 Modular Steam
Generators
Mechanical
pump
Intermediate circuit with
Mechanical Pump ALIP
Intermediate circuit with ALIP
ALIP vs Mechanical Pump
ASTRIDrsquos power conversion system
higher reliability no moving part
no leakage risk
simplification of the design
reduced maintenance
cost effective
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Comparison between mechanical and electromagnetic pumps
ELECTROMAGNETIC PUMPS
1048633 Advantages
bullno rotating mechanical pieces
bullvery limited maintenance
bullgreat reliability (for BR10 as exemple 170000 hrs(~20 years) of
operation without major incident (same for ancillary system of
SPX)
bullvery small impact of cavitation
1048633 Drawbacks
bulllow efficiency (maximum 40)
bullrisks of electromagnetic instabilities for large pumps
bullimportant component volume required for very large flowrates (ie
some tonss)
bullelectrical insulation and magnetic materials working at 550degC
bullno operational feedback from large pumps immersed in reactor
MECANICAL PUMPS
1048633 Advantages
bull large operational feedback from reactors
bullgood efficiency (70 agrave 80)
bull Important inertia when stopped
1048633 Drawbacks
bullseveral rotating elements
bulllimited life duration for hydrostatic bearings
bullnecessity to cool engines bearings
bullneacutecessity of periodical maintenance
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Energy Conversion System
Main goal to eliminate or mitigate the risk of Na-Water reaction
2012 choice of an option
ECS gas (nitrogen 100) ECS steam
Rankine with modular
SGU
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Commercial 1500 MWe reactor
(3600 MWth)
1 Reactor
6 IHX (Stainless Steel)
3 Primary Pumps in pool
6 DHX x 50 in pool
Primary Fuel Handling with
Rotating Plugs
Energy Conversion
6 Secondary Loops each
equipped with 6 Modular SG
(100 MWth) SodiumWater
Classical Steam Turbine
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
kgs1650Water Flow Rate 6 loops
bars185SG Steam Pressure
degC490SG Outlet Temperature
degC240SG Inlet Temperature
kgs2555Secondary Flow Rate 6 loops
degC525IHX Outlet Temperature
degC340IHX Inlet Temperature
kgs19 000Core Flow Rate
degC545Core Mean Outlet Temperature
degC395Core Inlet Temperature
UNITVALUEDATA
CP-ESFR ndash General Hypothesis
A proposal for a coherent plant architecture
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Innovative Steam Generators
Robust steam generators
(double tubes modular
improved instrumentation hellip)
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Shell and tubes SGU
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Shell and tubes SGU
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Intermediate heat exchanger Phenix
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Heat transfer several existing technologies
Several parameters to check prior to
choice of technology
- maximal pressure amp temperature
- Compacity
- Efficiency
- Reliability (Thermal behaviourhellip)
- Inspectability
- Reparability
- Modularity
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
ndash Shell and tubes heat exchanger
ndash Plate Stamped Heat Exchanger (PSHEs)
ndash Printed Circuit Heat Exchangers (PCHEs)
Shell and tubes
PCHE
(Printed Compact
Heat exchanger)
View of a brazed plate heat exchangers (courtesy of Alfa-Laval and Ciat)
PSHE
(Plate Stamped
Heat exchanger)
Plate heat exchanger
Hsup2X
(Hybrid
HeateXchanger)
Technical solutions considered for Nagas heat exchanger
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Heat exchanger design
Ex For pipes
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Comparison of various Heat Exchangers for Brayton ECS (He-N2)
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Na-Sc-CO2 Brayton cycle
H2O
RANKINE cycle
Supercritical CO2
BRAYTON cycle
Interest of Na-Sc-CO2 Brayton
cycle
Potential better efficiency
Better compactness of the turbine
Less consequences than Na H2O reaction
T
HT recuperator BT Reacutecupeacuterator
Flow Split
Junction
SF
By-pass compressor Main Compressor
SC
Density (kgm3)
Pressure (bar)
Temperature
(degC)
Na-CO2 heat exchanger
Courtesy of CEA
SMFR CO2 ECS
Courtesy of DOE
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Liquid metals
Molten salts
NaI Alternative
coolant
H2O steam
IHX-SGU integrated
intermediate loop
Innovative options
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Innovative options
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Concentration-temperature diagram
raa
001
01
1
10
100
1000
10000
100
130
160
190
220
250
280
310
340
370
400
430
460
490
520
550
580
Temperature degC
[O] ppm
[H] ppm
log [ ( )]
( )10 6 2502444 5
O ppmT K
log [ ( )] ( )10 6 467
3023H ppm
T K
Noden solubility law Wittingham solubility law
O and H solubilities are
negligible close to 978degC
Consequences Na can be
purified by Na cooling
leading to
crystallization of O and H
as Na2O and NaH
in a cold trap
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Basic principle of a cold trap
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Large components Handling operations
Handling cask for IHX PP
(prevents from irradiation amp
sodium contamination)
IHX
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Fuel handling system
bull Reactor refueling system provides the means of transporting storing and
handling for reactor core assemblies including fuel blanket control and
shielding elements
bull FHS have to fulfill the following tasks
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Fuel assemblies Handling operations
Option with External Vessel in Na for FA storage
Option with Internal Gas Vessel for FA storage
Main requirements
- Insure loadingunloading of fuel
assemblies (FA)
-Cool down the irradiated fuel
assemblies
- transfer FA to intermediate
gasNawater storage
- eliminate residual Na from FA
- take into account FA with Minor
Actinides
- be able to manage FA with fuel
clad failure
With a limited duration FA
Two main option for In primary
vessel handling
-Two rotating plugs + Fuel
Handling device+ transfer ramp
-Pantograph
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Innovative Fuel Handling System
gasket
Core
Option JSFR + SMFR
1 rotating plug Pantograph arm
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
bull A Whole Core Discharge is an exceptional event which can be considered necessary in view of a comprehensive reactor inspection
bull Not considered in normal outages plans WCD could direct choices on FHS
ndash Sodium route is the preferred solution for fast whole core discharge
ndash Duration of a WCD has to be about 1 to 3 months
bull Design of External Vessel Storage Tank
ndash Filled with sodium (400m3)
ndash 800 storage positions in less than 8 meters
ndash Total inspection is possible and all components are easy to maintain
bull Final decision concerning context of WCD will include other considerations such as global economy and safety optimizations
Fast whole core discharge External Vessel Storage Tank
External Vessel
Storage Tank
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Cleaning process (Na) before repairing (Phenix process)
Before cleaning After cleaning
Gas sweeping
CO2 andor Ar
Gas outlet
Spray nozzles
Injection CO2
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Conclusion
The development of Fast Neutron Reactors is essential for the future energy
supply
Na is today recognized as the best primary coolant for Fast Neutrons Reactors
SFR technology is rather well known but the operational feedback from existing
reactors has shown that reliability of components and systems has to be
improved The reactor availability is a key indicator of a mature concept
The major objective for the future SFR development is the improvement of the
economy and safety of the systems and more particularly the improvement of
performances (burn-up energy conversion system) simplification of systems
(intermediate loop handling systems componentshellip) improvement of the In
service inspection reparability operability and availability (fuel handling
repair duration)
Operational constraints have to be considered at the design stage Innovative
designs have to be evaluated also with regards operation thanks to operational
reactor feedback analysis and modeling from existing and future reactors
Collaborations are welcome to contribute the this main goal provide energy for
the future generations
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Thank You very much for your kind attention
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID
Power [MWth] [MWe] 1500 (~600)
Thermal efficiency 40 (for Rankine cycle) 37 (for Brayton cycle with nitrogen)
Core inlet-outlet temperature [degC] 400degC - 550degC
Fuel MOX considerations for carbides for future (with and without MA)
Reactor Vessel Hanged Austenitic stainless steel
Safety Vessel Anchored to reactor pit(or hanged not yet decided)
Fuel cladding temperature [degC] Maximum 700degC (permanent state)
Cladding material
Hexcan
15-15 Ti (AIM1) as reference ODS being investigated for future core
martensitic steel 9Cr EM10
Primary system Pool type compact forced circulation natural circulation for DHR
Primary system pressure loss lt 05 MPa (Unprotected Loss of Flow grace time gtx min)
Allowed maximum Na velocity (ms) No specification due to non significant ersosion-corrosion
Primary pumps Mechanical pumps
Number of intermediate loops 2 to 4 (depends on safety and economy)
Intermediate pumps Mechanical pumps or Electromagnetic pumps
Energy Conversion System Water-superheated steam at 180 bar in modular Steam Generator
Pure nitrogen at 180 bar in modular Na-gas heat exchanger
Reactor Internals Ability to be inspected (Whole Core discharge possible)
Seismic design provisions Classical systems for Sodium Fast Reactors
Number of shutdown systems 2 and a third innovative device under investigation
DHR systems Several architectures investigated
Severe accidents Recuperator for corium (internal or external)
Main parameters of ASTRID