Introduction to Generation IV
Nuclear Energy Systems
Dr. Ralph Bennett, Technical Director, Generation IV International Forum, and
Director, International and Regional Partnerships, Idaho National Laboratory
16 Mar 2009
The Problem of Climate Change
•
•
•
Global greenhouse gas (GHG) emissions have grown since
pre-industrial times, increasing 70% between 1970 and 2004
With current climate change mitigation policies and practices,
global GHG emissions will continue to grow
The Earth is about to undergo long lasting changes in its
climate, seas and land cover, including
– Temperature
Global Warming (deg C) by 2100 (IPCC prediction)
– Precipitation
–
–
–
–
–
Sea level
Ocean circulation
Ice/snow cover
Storm frequency
Storm intensity
–
Desertification
http://www.aip.org/history/climate/summary.htm
http://www.grida.no/climate/ipcc_tar/
2
The Challenge for Nuclear Energy
•
•
Nuclear is a major contributor in the WEO 2008 450 Policy Scenario—
about 250 GWe more generation by 2030 (an 80% increase from today)
Nuclear energy systems must continue their advances in order to
unlock a potential on this scale
http://www.iea.org/weo/2008.asp
3
Generations of Nuclear Energy
Generation IV
Generation III+
Generation III
Generation I
Early Prototypes
- Shippingport
- Dresden
- Magnox
1950
1960
Gen I
Revolutionary
Designs
Evolutionary Designs
Generation II
Advanced LWRs
Commercial Power
1970
- ABWR
- ACR1000
- AP1000
- APWR
- EPR
- ESBWR
- CANDU 6
- System 80+
- AP600
- PWRs
- BWRs
- CANDU
1980
Gen II
1990
- Safe
- Sustainable
- Economical
- Proliferation
2000
Gen III
2010
Resistant and
Physically
Secure
2020
Gen III+
2030
Gen IV
http://www.gen-4.org/Technology/evolution.htm
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Creation of the International Forum
• Started in Jan 2000 by nine countries and established Jul 2001
• Agreed that nuclear energy is needed to meet future needs
• Defined four goal areas to advance nuclear energy into its
next, ‘fourth’ generation:
– Sustainability
– Safety & reliability
– Economics
– Proliferation resistance and physical protection
• Will collaborate to make ‘Generation IV’ systems deployable in
large numbers by 2030, or earlier
http://www.gen-4.org/GIF/About/origins.htm
5
Today’s Membership
6
Overview of the Generation IV Systems
Neutron
Spectrum
Fuel
Cycle
Size
(MWe)
Sodium Cooled Fast
Reactor (SFR)
Fast
Closed
Very-HighTemperature Reactor
(VHTR)
Thermal
Gas-Cooled Fast
Reactor (GFR)
System
Missions
R&D Needed
300-1500
Electricity, Actinide
Management
Advanced recycle
options, Fuels
Open
250
Electricity, Hydrogen,
Process Heat
Fuels, Materials,
H2 production
Fast
Closed
1200
Electricity, Hydrogen,
Actinide Management
Fuels, Materials,
Thermal-hydraulics
Supercritical-Water
Reactor (SCWR)
Thermal,
Fast
Open,
Closed
1500
Electricity
Materials, Thermalhydraulics
Lead-Cooled Fast
Reactor (LFR)
Fast
Closed
50-150
300-600
1200
Electricity,
Hydrogen Production
Fuels, Materials
Molten Salt Reactor
(MSR)
Epithermal
or Fast
Closed
1000
Electricity, Hydrogen
Production, Actinide
Management
Fuel treatment,
Materials, Reliability
http://www.gen-4.org/Technology/systems/index.htm
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Sodium-Cooled Fast Reactor (SFR)
Characteristics
• Sodium coolant, pool or loop type
• 550C outlet temperature
• 600-1500 MWe large size, or
• 300-600 MWe intermediate size
• 50 MWe small module option
• Metal fuel with pyroprocessing or
MOX fuel with advanced aqueous
separation
Benefits
• High thermal efficiency
• Consumption of LWR actinides
• Efficient fissile material
generation
http://www.gen-4.org/Technology/systems/index.htm
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SFR Reactor Options
Large-scale Loop
AHX
Chimney
Intermediate-scale Pool
Secondary Pump
PDRC
piping
IHTS
piping
Steam
Generator
SG
Primary
Pump/IHX
Small-scale Modular
CONTROL
BUILDING
IHX
DHX
PHTS
pump
Reactor
core
TURBINE/GENERATOR
BUILDING
Na-Air
HEAT EXCHANGER (2)
IHTS
pump
ELEVATOR
Ø 7.7m
(Ø 25.5')
In-vessel core
catcher
EXHAUST TO VENT STACK
Na-CO2
HEAT EXCHANGER
.61m
[2FT]
3
12.03 m
3,186 gal.
Hot Pool
Normal sodium level
Reactor Vessel
1.93m
[6.3FT]
IHX
X-SECTION (FLATTENED FOR CLARITY)
IHX
14.76m
[48.4FT]
PUMPS (2)
ON Ø 142.5" B.C.
12.72m
[41.7FT]
3.25m
(10'-8")
PRIMARY
CONTROL RODS
SECONDARY
CONTROL RODS
CORE BARREL Ø
266 / 268 cm
(104.7" / 105.5")
0.75m
(29.5")
0
1
2
3
4
5
METERS
GUARD VESSEL
(1" THICK)
THERMAL
SHIELD
Primary Vessel I.D.
Guard Vessel I.D.
IHX (2)
1.7m2 EACH
3.5m
1m
(11'-8")
(39.4")
PRIMARY VESSEL
(2" THICK)
1m TRAVEL DISTANCE
OF THE CONTROL RODS
PLAN VIEW OF THE CORE
DRACS (2)
0.4m2 EACH
PLAN VIEW OF
IHX AND PUMPS
2.29m
[7.5FT]
Sodium faulted level
7m
[23FT]
SODIUM DUMP TANK
Ø 2.5 m x 3.8 m LONG
(Ø 7.5' x 12.6' LONG)
CONTROL
RODS (7)
Pump off
Sodium Level
Cold Pool
Normal sodium level
1.89m
[6.2FT]
7m
(23')
10
4.57m
[15FT]
5.08m
[16.7FT]
SECTION A - A
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SFR Technology Interests
•
Minor actinide bearing fuel technology (fabrication, irradiation)
– Metal and oxide fuel performance
– Carbide fuel performance
– Nitride/Carbide fuel performance
•
Inspection & repair technologies
– Ultrasonic and alternative techniques
– Replace/repair experience
•
High temperature leak-before-break assessment technologies
– Creep-fatigue crack initiation and growth test results
•
Advanced energy conversion concepts
– Basic design concept of supercritical CO2 Brayton cycle system
– Compact supercritical CO2-to-CO2 heat exchangers
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Very-High-Temperature Reactor (VHTR)
Characteristics
• He coolant
• >900C outlet temperature
• 250 MWe
• Coated particle fuel in either
pebble bed or prismatic fuel
Benefits
• Hydrogen production
• Process heat applications
• High degree of passive safety
• High thermal efficiency option
http://www.gen-4.org/Technology/systems/index.htm 11
VHTR Reactor Options
Pebble bed core
Prismatic-fuel core
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VHTR Hydrogen Options
Sulfur-iodine cycle
High temperature electrolysis
90 v/o H2 O + 10 v/o H2
4 e-
10 v/o H2O + 90 v/o H2
Porous Cathode, Nickel -Zirconia cermet
H2O


H2
2 H20 + 4 e -  2 H2 + 2 O =
2 O=

Gastight Electrolyte, Yttria-Stabilized Zirconia
2 O =  O2 + 4 e -
O2

Porous Anode, Strontium -doped Lanthanum Manganite
 
Interconnection
H2O + H2 
H2O

Next Nickel-Zirconia Cermet Cathode

H2
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VHTR Technology Interests
•
Fuel and fuel cycle
– Particle fuel irradiations and fission product monitoring
•
Materials
– Codes and standards extension
– Materials database extension
– Graphite dust behavior
•
Hydrogen production
–
–
–
–
•
•
Sulfur-iodine cycle
High temperature electrolysis
Coupling of H2 production process and reactor heat transport system
Tritium transport
Computational Methods
Components and helium turbine
– Intermediate heat exchanger
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Lead-Cooled Fast Reactor (LFR)
Characteristics
• Pb or Pb/Bi coolant
• 550C to 800C outlet temperature
• Small transportable system 50150 MWe, and
• Larger station 300-1200 MWe
• 15–30 year core life option
Benefits
• Distributed electricity generation
• Hydrogen and potable water
• Replaceable core for regional
fuel processing
• High degree of passive safety
• Proliferation resistance through
long-life core
http://www.gen-4.org/Technology/systems/index.htm 15
LFR Reactor Options
Small, transportable module
CLOSURE HEAD
CO2 OUTLET NOZZLE
(1 OF 8)
CO 2 INLET NOZZLE
(1 OF 4)
Pb-TO-CO 2 HEAT
EXCHANGER (1 OF 4)
FLOW SHROUD
RADIAL REFLECTOR
Large, stationary plant
CONTROL
ROD
DRIVES
CONTROL
ROD GUIDE
TUBES AND
DRIVELINES
THERMAL
BAFFLE
GUARD
VESSEL
REACTOR
VESSEL
ACTIVE CORE AND
FISSION GAS PLENUM
FLOW DISTRIBUTOR
HEAD
• Pb coolant (both)
• No intermediate loops
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LFR Technology Interests
•
Collaborations based on ELSY and SSTAR
– No formal agreement yet
•
Conceptual design and safety
– Innovative components and design
– Compact, in-vessel steam generators
– Decay heat removal by air and water
– Refueling ‘out-of-Pb’ coolant
– Innovative structural design
– Buoyant fuel element support
– Seismic isolation of reactor building
•
Fuel and core materials
– Many options
ELSY: European Lead-cooled System; SSTAR: Small Secure Transportable Autonomous Reactor
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Supercritical-Water-Cooled Reactor (SCWR)
Characteristics
• Water coolant above
supercritical conditions
(374C, 22.1 MPa)
• 510-625C outlet temperature
• 1500 MWe
• Pressure tube or pressure
vessel options
• Simplified balance of plant
Benefits
• Efficiency near 45% with
excellent economics
• Leverages the current
experience in operating fossilfueled supercritical steam
plants
• Configurable as a fast- or
thermal-spectrum core
http://www.gen-4.org/Technology/systems/index.htm 18
Gas-Cooled Fast Reactor (GFR)
Characteristics
• He coolant
• 850C outlet temperature
• Direct gas-turbine cycle or
supercritical CO2 cycle with
optional combined cycles
• 2400 MWth / 1100 MWe
• Several fuel options
– Carbide in plates or pins
– Nitride
– Oxide
Benefits
• High efficiency
• Waste minimization and
efficient use of uranium
resources
http://www.gen-4.org/Technology/systems/index.htm 19
Molten Salt Reactor (MSR)
Characteristics
• Fuel is liquid fluorides of U or Th
with Li, Be, Na and other fluorides
• 700–800C outlet temperature
• 1000 MWe
• Low pressure (<0.5 MPa)
Benefits
• Waste minimization
• Avoids fuel development
• Proliferation resistance through
low fissile material inventory
http://www.gen-4.org/Technology/systems/index.htm 20
Organization
Policy Group
Senior Industry
Advisory Panel
Chair (France)
Experts Group
Policy Secretariat
Chair
Policy
Director
Technical
Director
System Steering
Committees
Co-Chairs
Methodology
Working Groups
Technical Secretariat
Proliferation Resistance
and Physical Protection,
Risk & Safety, Economics
NEA, Paris
Project Management
Boards
(multiple R&D projects)
http://www.gen-4.org/GIF/Governance/index.htm
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System Partners
Mar 2009
VHTR

GFR
SFR
SCWR

LFR
MSR
Partners:
ANRE
CAEA
CEA
DME
DOE
JAEA
JRC
KOSEF
MEST
MOST
NRCan
PSI
NRCan

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





JRC
CEA


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JAEA,
ANRE
MEST,
KOSEF

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

PSI
DOE
CAEA,
MOST


– Agency for Natural Resources and Energy (JP)
– China Atomic Energy Authority (CN)
– Commissariat à l’Énergie Atomique (FR)
– Department of Minerals and Energy (ZA)
– Department of Energy (US)
– Japan Atomic Energy Agency (JP)
– Joint Research Centre (EU)
– Korean Science and Engineering Foundation (KR)
– Ministry of Education, Science and Technology (KR)
– Ministry of Science and Technology (CN)
– Natural Resources Canada (CA)
– Paul Scherrer Institute (CH)
VHTR
GFR
SFR
SCWR
LFR
MSR
DME
– Very-High-Temperature Reactor
– Gas-Cooled Fast Reactor
– Sodium-Cooled Fast Reactor
– Supercritical Water-Cooled Reactor
– Lead-Cooled Fast Reactor
– Molten Salt Reactor
http://www.gen-4.org/GIF/Governance/system.htm
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Generation IV Annual Report
•
Captures key information and accomplishments
from System Steering Committee annual reports
into one widely distributed report
•
Captures brief summaries of working groups’
accomplishments, and background on the Forum
•
Audience includes:
– World-wide Research and Development
Community
– Governments sponsoring Generation IV R&D
– GIF committees, boards and working groups
•
The 2008 Report has just issued
http://www.gen-4.org/PDFs/GIF_2008_Annual_Report.pdf
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Working Toward the Future
The GIF joined together to help assure a sustainable energy future
• Underscored by the advance of global climate change
• Based on advanced nuclear energy systems that are sustainable,
safe, economical, proliferation resistant and physically secure
• Accelerated by the collaboration of the GIF members, industry,
academia and non-member nations and institutions
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Bibliography
• The web links provided on most slides lead to source
documents, background materials or updates
• The full Generation IV Roadmap and all supporting documents
are available at: http://gif.inel.gov/roadmap/
• Some technical papers are listed on the OECD NEA website (GIF
website) at www.gen-4.org within each system
• Recent outlook articles on nuclear deployment:
–
–
–
–
–
–
IEA http://www.iea.org/weo/2008.asp (subscription)
NEA http://www.nea.fr/neo/ (subscription)
IAEA http://www-pub.iaea.org/MTCD/publications/PDF/RDS1-28_web.pdf
WNA http://www.world-nuclear.org/outlook/clean_energy_need.html
EPRI (US R&D strategy and deployment outlook, respectively)
http://my.epri.com/portal/server.pt?Product_id=000000000001018514.pdf
http://my.epri.com/portal/server.pt?Product_id=000000000001018431.pdf
• My contact information:
–
ralph.bennett@inl.gov
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