Sodium Fast Reactors Systems and components (Part 2) · PDF file ·...

32
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 : [email protected] IAEA Education &Training Seminar on Fast Reactor Science and Technology CNEA Bariloche, Argentina October 1 5, 2012 Sodium Fast Reactors Systems and components (Part 2)

Transcript of Sodium Fast Reactors Systems and components (Part 2) · PDF file ·...

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