Generation IV Nuclear Reactors
(updated July 2013)
http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/Power-Reactors/Generation-IV-Nuclear-Reactors/#.UhhcJpFgw-I
- An international task force is developing six nuclear reactor technologies for deployment between 2020 and 2030. Four are fast neutron reactors.
- All of these operate at higher temperatures than today's reactors. In particular, four are designated for hydrogen production.
- All six systems represent advances in sustainability, economics, safety, reliability and proliferation-resistance.
- Europe is pushing ahead with three of the fast reactor designs.
GIF
The Generation IV International Forum
(GIF) was initiated in 2000 and formally chartered in mid 2001. It is
an international collective representing governments of 13 countries
where nuclear energy is significant now and also seen as vital for the
future. Most are committed to joint development of the next generation
of nuclear technology. Led by the USA,
Argentina, Brazil, Canada, China, France, Japan, Russia, South Korea,
South Africa, Switzerland, and the UK are charter members of the GIF,
along with the EU (Euratom). Most of these are party to the Framework
Agreement (FA) which formally commits them to participate in the
development of one or more Generation IV systems selected by GIF for
further R&D. Argentina and Brazil did not sign the FA, and the UK
withdrew from it; accordingly, within the GIF, these three are
designated as “inactive Members.” Russia formalized its accession to the
FA in August 2009 as its tenth member, with Rosatom as implementing
agent. In 2011 the 13 members decided to modify and extend the GIF
charter indefinitely.
GIF focus
After some two years' deliberation and review of about one hundred
concepts, GIF (then representing ten countries) late in 2002 announced
the selection of six reactor technologies which they believe represent
the future shape of nuclear energy. These were selected on the basis of
being clean, safe and cost-effective means of meeting increased energy
demands on a sustainable basis, while being resistant to diversion of
materials for weapons proliferation and secure from terrorist attacks.
They are the subject of further development internationally, and
expenditure so far is in line with the initial estimate of $6 billion
over 15 years. About 80% of the cost is being met by the USA, Japan and
France.
In addition to selecting these six concepts for deployment between
2010 and 2030, the GIF recognised a number of International Near-Term
Deployment advanced reactors available before 2015. (see Advanced Reactors paper )
Most of the six systems employ a closed fuel cycle to maximise the
resource base and minimise high-level wastes to be sent to a repository.
Three of the six are fast neutron reactors (FNR)
and one can be built as a fast reactor, one is described as epithermal,
and only two operate with slow neutrons like today's plants.
Only one is cooled by light water, two are helium-cooled and the
others have lead-bismuth, sodium or fluoride salt coolant. The latter
three operate at low pressure, with significant safety advantage. The
last has the uranium fuel dissolved in the circulating coolant.
Temperatures range from 510°C to 1000°C, compared with less than 330°C
for today's light water reactors, and this means that four of them can
be used for thermochemical hydrogen production.
The sizes range from 150 to 1500 MWe (or equivalent thermal) , with
the lead-cooled one optionally available as a 50-150 MWe "battery" with
long core life (15-20 years without refuelling) as replaceable cassette
or entire reactor module. This is designed for distributed generation or
desalination.
At least four of the systems have significant operating experience
already in most respects of their design, which provides a good basis
for further R&D and is likely to mean that they can be in commercial
operation well before 2030.
However, it is significant that to address non-proliferation
concerns, the fast neutron reactors are not conventional fast breeders,
ie they do not have a blanket assembly where plutonium-239 is produced.
Instead, plutonium production takes place in the core, where burn-up is
high and the proportion of plutonium isotopes other than Pu-239 remains
high. In addition, new reprocessing technologies will enable the fuel
to be recycled without separating the plutonium.
In February 2005 five of the participants signed an agreement to take
forward the R&D on the six technologies. The USA, Canada, France,
Japan and UK agreed to undertake joint research and exchange technical
information.
Associated ongoing programs
While Russia was not initially part of GIF, one design corresponds
with the BREST reactor being developed there, and Russia is now the main
operator of the sodium-cooled fast reactor for electricity - another of
the technologies put forward by the GIF.
India is also not involved with the GIF but is developing its own
advanced technology to utilise thorium as a nuclear fuel. A three-stage
program has the first stage well-established, with Pressurised Heavy
Water Reactors (PHWRs, elsewhere known as CANDUs) fuelled by natural
uranium to generate plutonium. Then Fast Breeder Reactors (FBRs) use
this plutonium-based fuel to breed U-233 from thorium, and finally
advanced nuclear power systems will use the U-233. The spent fuel will
be reprocessed to recover fissile materials for recycling. The two major
options for the third stage, while continuing with the PHWR and FBR
programs, are an Advanced Heavy Water Reactor and subcritical
Accelerator-Driven Systems.
Closely related to GIF is the Multinational Design Evaluation Program(MDEP)
set up in 2005, led by the OECD Nuclear Energy Agency and involving the
IAEA. It aims to develop multinational regulatory standards for design
of Gen IV reactors. The US Nuclear Regulatory Commission (NRC) has
proposed a three-stage process culminating in international design
certification for these. Ten countries are involved so far: Canada,
China, Finland, France, Japan, Korea, Russia, South Africa, UK, USA, but
others which have or are likely to have firm commitments to building
new nuclear plants may be admitted. In September 2007 the NRC called for
countries involved in development of Gen IV reactors to move to stage 3
of design evaluation, which means developing common design requirements
so that regulatory standards can be harmonised. NRC has published its
draft design requirements.
A major project is investigating the use of actinide-laden fuel
assemblies in fast reactors as part of the sodium-cooled fast reactor
program. The Global Actinide Cycle International Demonstration (GACID)
is being undertaken by France's atomic energy commission (CEA), Japan's
Atomic Energy Agency (JAEA) and the US Department of Energy (DOE) under
the US Advanced Fuel Cycle Initiative (AFCI). The first stage will lead
to demonstration fuel containing minor actinides being used in Japan's
Monju reactor.
GIF Reactor technologies:
There were originally six technologies chosen, but development on
one has gone in two directions, so seven are listed in the Table below.
Gas-cooled fast reactors.
Like other helium-cooled reactors which have operated or are under
development, GFRs will be high-temperature units - 850°C. They employ
similar reactor technology to the VHTR, suitable for power generation,
thermochemical hydrogen production or other process heat. The reference
GFR unit is 1200 MWe, with thick steel reactor pressure vessel and
three 800 MWt loops. For electricity, the helium will directly drive a
gas turbine (Brayton cycle). It would have a self-generating (breeding)
core with fast neutron spectrum and no fertile blanket. Robust nitride
or carbide fuels would include depleted uranium and any other fissile
or fertile materials as ceramic pins or plates, with plutonium content
of 15 to 20%. As with the SFR, used fuel would be reprocessed on site
and all the actinides recycled repeatedly to minimise production of
long-lived radioactive wastes.
While General Atomics worked on the design in the 1970s (but not as
fast reactor), none has so far been built. It is the only Gen IV design
with no operating antecedent, so a prototype is not expected before
2025. However, an 80 MWt experimental technology demonstration GFR,
ETDR or ALLEGRO, is planned by Euratom to be built from 2014. It will
incorporate all the architecture and the main materials and components
foreseen for the GFR without the power conversion system. Euratom,
France, Japan and Switzerland have signed on to System Arrangements (SA)
for the GFR under the Framework Agreement. See also European program
section below.
An alternative GFR design has lower temperature (600-650ºC) helium
cooling in a primary circuit and supercritical CO2 at 550ºC and 20 MPa
in a secondary system for power generation. This reduces the
metallurgical and fuel challenges associated with very high
temperatures.
Lead-cooled fast reactors.
The LFR is a flexible fast neutron reactor which can use depleted
uranium or thorium fuel matrices, and burn actinides from LWR fuel.
Liquid metal (Pb or Pb-Bi eutectic) cooling is at low pressure by
natural convection (at least for decay heat removal). Fuel is metal or
nitride, with full actinide recycle from regional or central
reprocessing plants. A wide range of unit sizes is envisaged, from
factory-built "battery" with 15-20 year life for small grids or
developing countries, to modular 300-400 MWe units and large single
plants of 1400 MWe. Operating temperature of 550°C is readily achievable
but 800°C is envisaged with advanced materials to provide lead
corrosion resistance at high temperatures and this would enable
thermochemical hydrogen production. A two-stage development program
leading to industrial deployment is envisaged: by 2025 for reactors
operating with relatively low temperature and power density, and by 2035
for more advanced higher-temperature designs.
This corresponds with Russia's BREST fast reactor technology which is
lead-cooled and builds on 80 reactor-years experience of lead or
lead-bismuth cooling, mostly in submarine reactors. Its fuel is U+Pu
nitride. More immediately the GIF proposal appears to arise from two
experimental designs: the US STAR and Japan's LSPR, these being lead and
lead-bismuth cooled respectively.
Initial development work on the LFR is focused on two pool-type
reactors: SSTAR - Small Secure Transportable Autonomous Reactor of 20
MWe in USA and the European Lead-cooled SYstem (ELSY) of 600 MWe in
Europe.
SSTAR is being developed by Toshiba and others in Japan. It runs at
566°C and has integral steam generator inside the sealed unit, which
would be installed below ground level. It is expected to have 44%
thermal efficiency. After a 20-year life without refuelling, the whole
reactor unit is then returned for recycling the fuel. The core is one
metre high and 1.2 m diameter (20 MWe version). SSTAR will eventually
be coupled to a Brayton cycle turbine using supercritical carbon dioxide
with natural circulation to four heat exchangers.
The ELSY project is led by Ansaldo Nucleare from Italy and is being
financed by Euratom. The 600 MWe design was nearly complete in 2008 and
a small-scale demonstration facility is planned. It runs on MOX fuel
at 480°C and the molten lead is pumped to eight steam generators.
For the LFR, no System Arrangements (SA) have been signed, and
collaborative R&D is pursued by interested members under the
auspices of a provisional steering committee. In 2011 Russia joined
them. A technology pilot plant is envisaged in operation by 2020,
followed by a prototype of a large unit and deployment of small
transportable units. See also European program section below.
Molten salt reactors (now two variants).
In an MSR, the uranium fuel is dissolved in the sodium fluoride salt
coolant which circulates through graphite core channels to achieve some
moderation and an epithermal neutron spectrum. The reference plant is up
to 1000 MWe. Fission products are removed continuously and the
actinides are fully recycled, while plutonium and other actinides can be
added along with U-238, without the need for fuel fabrication. Coolant
temperature is 700°C at very low pressure, with 800°C envisaged. A
secondary coolant system is used for electricity generation, and
thermochemical hydrogen production is also feasible.
Compared with solid-fuelled reactors, MSR systems have lower fissile
inventories, no radiation damage constraint on fuel burn-up, no spent
nuclear fuel, no requirement to fabricate and handle solid fuel, and a
homogeneous isotopic composition of fuel in the reactor. These and
other characteristics may enable MSRs to have unique capabilities and
competitive economics for actinide burning and extending fuel resources.
During the 1960s the USA developed the molten salt fast reactor as
the primary back-up option for the conventional fast breeder reactor,
and a small prototype was operated for about four years. Recent work has
focused on lithium and beryllium fluoride coolant with dissolved
thorium and U-233 fuel. The attractive features of the MSR fuel cycle
include: the high-level waste comprising fission products only, hence
shorter-lived radioactivity; small inventory of weapons-fissile material
(Pu-242 being the dominant Pu isotope); low fuel use (the French
self-breeding variant claims 50kg of thorium and 50kg U-238 per billion
kWh); and safety due to passive cooling up to any size.
For the MSR, no System Arrangements (SA) have been signed, and
collaborative R&D is pursued by interested members under the
auspices of a provisional steering committee. There will be a long lead
time to prototypes, and the R&D orientation has changed since the
project was set up, due to increased interest. It now has two baseline
concepts:
- the Molten Salt Fast Neutron Reactor (MSFR)
- the Advanced High-Temperature Reactor (AHTR) with the same graphite core structures as the VHTR and molten salt as coolant instead of helium, enabling power densities 4 to 6 times greater than HTRs and power levels up to 4000 MWt with passive safety systems.
- the Molten Salt Fast Neutron Reactor (MSFR)
- the Advanced High-Temperature Reactor (AHTR) with the same graphite core structures as the VHTR and molten salt as coolant instead of helium, enabling power densities 4 to 6 times greater than HTRs and power levels up to 4000 MWt with passive safety systems.
Sodium-cooled fast reactors .
The SFR uses liquid sodium as the reactor coolant, allowing high power
density with low coolant volume. It builds on some 390 reactor-years
experienced with sodium-cooled fast neutron reactors over five decades
and in eight countries, and is the main technology of interest in GIF.
Most plants so far have had a core plus blanket configuration, but new
designs are likely to have all the neutron action in the core. Other
R&D is focused on safety in loss of coolant scenarios, and improved
fuel handling.
The SFR utilises depleted uranium as the fuel matrix and has a
coolant temperature of 500-550°C enabling electricity generation via a
secondary sodium circuit, the primary one being at near atmospheric
pressure. Three variants are proposed: a 50-150 MWe type with actinides
incorporated into a U-Pu metal fuel requiring electrometallurgical
processing (pyroprocessing) integrated on site, a 300-1500 MWe pool-type
version of this, and a 600-1500 MWe type with conventional MOX fuel and
advanced aqueous reprocessing in central facilities elsewhere.
Early in 2008, the USA, France and Japan signed an agreement to
expand their cooperation on the development of sodium-cooled fast
reactor technology. The agreement relates to their collaboration in the
Global Nuclear Energy Partnership, aimed at closing the nuclear fuel
cycle through the use of advanced reprocessing and fast reactor
technologies, and seeks to avoid duplication of effort.
Euratom, China, France, Japan, Korea and the USA have signed on to
System Arrangements (SA) for the SFR under the Framework Agreement, and
in 2011 Russia joined them. Three Project Arrangements have been signed
within the SFR system: the Advanced Fuel PA; the Global Actinide Cycle
International Demonstration (GACID) PA; and the Component Design and
Balance-Of-Plant PA. See also European program section below.
Supercritical water-cooled reactors .
This is a very high-pressure water-cooled reactor which operates above
the thermodynamic critical point of water (374ºC, 22 MPa) to give a
thermal efficiency about one third higher than today's light water
reactors from which the design evolves. The supercritical water (25 MPa
and 510-550°C) directly drives the turbine, without any secondary steam
system,* simplifying the plant. Two design options are considered:
pressure vessel and pressure tube. Passive safety features are similar
to those of simplified boiling water reactors. Fuel is uranium oxide,
enriched in the case of the open fuel cycle option. However, it can be
built as a fast reactor with full actinide recycle based on conventional
reprocessing.
Euratom, Canada and Japan have signed on to System Arrangements (SA)
for the SCWR under the Framework Agreement. Project Arrangements are
pending for thermal-hydraulics and safety. Pre-conceptual SCWR designs
include Candu (Canada), LWR (Euratom) and Fast Neutron (Japan).
* Today's supercritical coal-fired plants use
supercritical water around 25 MPa which have "steam" temperatures of 500
to 600ºC and can give 45% thermal efficiency. At ultra supercritical
levels (30+ MPa), 50% thermal efficiency may be attained. Over 400 such
plants are operating world-wide.
Supercritical fluids are those above the thermodynamic
critical point, defined as the highest temperature and pressure at
which gas and liquid phases can co-exist in equilibrium. They have
properties between those of gas and liquid. For water the critical
point is at 374°C and 22 MPa, giving it a "steam" density one third that
of the liquid so that it can drive a turbine in a similar way to normal
steam.
Very high-temperature gas reactors .
These are graphite-moderated, helium-cooled reactors, based on
substantial experience.Euratom, Canada and Japan have signed on to
System Arrangements (SA) for the SCWR under the Framework Agreement.
Project Arrangements are pending for thermal-hydraulics and safety.
Pre-conceptual SCWR designs include Candu (Canada), LWR (Euratom) and
Fast Neutron (Japan).
The core can be built of prismatic blocks such as the Japanese HTTR
and the GTMHR under development by General Atomics and others in Russia,
or it may be pebble bed such as the Chinese HTR-10 or HTR-PM and the
PBMR under development in South Africa, with international partners.
Outlet temperature of over 900°C and aiming for 1000ºC enables
thermochemical hydrogen production via an intermediate heat exchanger,
with electricity cogeneration, or direct high-efficiency driving of a
gas turbine (Brayton cycle). There is some flexibility in fuels, but no
recycle initially. Modules of 600 MW thermal are envisaged. The VHTR
has potential for high burn-up (150-200 GWd/t), completely passive
safety, low operation and maintenance costs, and modular construction.
Euratom, Canada, France, Japan, China, Korea, Switzerland and the USA
have signed on to the System Arrangement (SA) for the VHTR under the
Framework Agreement. South Africa is expected to do so in 2009. Two
Project Arrangements have been signed within the VHTR system: the Fuel
and Fuel Cycle PA and the Hydrogen Production PA. A Materials PA is
pending and will involve PBMR Pty Ltd.
neutron spectrum (fast/ thermal) |
coolant | temperature (°C) |
pressure* | fuel | fuel cycle | size(s) (MWe) |
uses | |
Gas-cooled fast reactors |
fast
|
helium
|
850
|
high
|
U-238 +
|
closed, on site
|
1200
|
electricity
& hydrogen |
---|---|---|---|---|---|---|---|---|
Lead-cooled fast reactors |
fast
|
lead or Pb-Bi
|
480-800
|
low
|
U-238 +
|
closed, regional
|
20-180**
300-1200 600-1000 |
electricity
& hydrogen |
Molten salt fast reactors |
fast
|
fluoride salts
|
700-800
|
low
|
UF in salt
|
closed
|
1000
|
electricity
& hydrogen |
Molten salt reactor - Advanced High-temperature reactors | thermal | fluoride salts | 750-1000 | UO2 particles in prism | open | 1000-1500 | hydrogen | |
Sodium-cooled fast reactors |
fast
|
sodium
|
550
|
low
|
U-238 & MOX
|
closed
|
30-150
300-1500 1000-2000 |
electricity
|
Supercritical water-cooled reactors |
thermal or fast
|
water
|
510-625
|
very high
|
UO2
|
open (thermal)
closed (fast) |
300-700
1000-1500 |
electricity
|
Very high temperature gas reactors |
thermal
|
helium
|
900-1000
|
high
|
UO2
prism or pebbles |
open
|
250-300
|
hydrogen
& electricity |
* high = 7-15 Mpa
+ = with some U-235 or Pu-239
** 'battery' model with long cassette core life (15-20 yr) or replaceable reactor module.
+ = with some U-235 or Pu-239
** 'battery' model with long cassette core life (15-20 yr) or replaceable reactor module.
European program from 2010
The European Commission in 2010 launched the European Sustainable
Nuclear Industrial Initiative (ESNII), which will support three
Generation IV fast reactor reactor projects as part of the EU’s plan to
promote low-carbon energy technologies. Other initiatives supporting
biomass, wind, solar, electricity grids and carbon sequestration are in
parallel. ESNII will take forward: the Astrid sodium-cooled fast reactor
(SFR) proposed by France, The Allegro gas-cooled fast reactor (GFR)
supported by central and eastern Europe, and the Myrrha lead- cooled
fast reactor (LFR) technology pilot proposed by Belgium.
The aim of ESNII is to demonstrate Gen IV reactor technologies that
can close the nuclear fuel cycle, provide long-term waste management
solutions, and expand the applications of nuclear fission beyond
electricity production to hydrogen production, industrial heat and
desalination. ESNII is designed to combine European capabilities in fast
neutron reactor R&D with industrial capability to build the
prototypes and develop supporting infrastructure.
The total estimated cost to ESNII of deploying these Gen IV
prototypes past 2020 is EUR 10.8 billion: EUR 5 billion for Astrid, EUR
1.96 billion is for Myrrha, a technology pilot and a later LFR
demonstrator, and EUR 1.2 billion for Allegro. Supporting infrastructure
is projected to cost EUR 2.65 billion. The 2010-12 ESNII budget is EUR
527 million, including EUR 329 million for Astrid.
Astrid SFR is led by the French CEA, involves EdF and Areva, and is
supported by a French government loan of EUR 651 million. Astrid is
based on about 45 reactor-years of operational experience in France and
will be rated 250 to 600 MWe. It is expected to be built at Marcoule
from 2017, with the unit being connected to the grid in 2022.
Allegro GFR is to be built in eastern Europe, and is more innovative.
It is rated at 100 MWt and would lead to a larger industrial
demonstration unit called GoFastR. The Czech Republic, Hungary and
Slovakia are making a joint proposal to host the project, with French
CEA support. Allegro is expected to begin construction in 2018 operate
from 2025. The industrial demonstrator would follow it.
In mid 2013 four nuclear research institutes and engineering
companies from central Europe’s Visegrád Group of Nations (V4) agreed
to establish a centre for joint research, development and innovation in
Generation IV nuclear reactors. The V4G4 Centre of Excellence is being
set up by scientific and research engineering company ÚJV Řež AS of the
Czech Republic, the Academy of Sciences Centre for Energy Research of
Hungary, Poland’s National Centre for Nuclear Research, and engineering
company VUJE AS of Slovakia. It is focused on gas-cooled fast reactors
such as Allegro.
Myrrha LFR project is initially a 57 MWt accelerator-driven system
with a liquid lead-bismuth (Pb-Bi) spallation target that in turn
couples to a Pb-Bi cooled, subcritical fast nuclear core. Later it will
become a European fast neutron technology pilot plant for lead and a
multi-purpose research reactor. Belgium’s SCK-CEN is leading the project
and will provide a total of about EUR 450 million. The unit is rated at
100 thermal MW and will be built at SCK-CEN’s Mol site beginning in
2014, and is to begin operation in 2023. A reduced-power model of Myrrha
called Guinevere started up at Mol in March 2010. ESNII also includes
an LFR technology demonstrator known as Alfred, also about 100 MWt, seen
as a prelude to an industrial demonstration unit of about 600 MWe.
Construction on Alfred could begin in 2017 and the unit could start
operating in 2025.
Sources:
US Department of Energy
DOE EIA 2003 New Reactor Designs.
US Department of Energy
DOE EIA 2003 New Reactor Designs.
GIF Annual Report 2008
http://www.gen-4.org/GIF/About/faq/index.htm
- I've heard a lot about Generation IV reactors. What are they and what are Generation II and III reactors?
- So, in what ways will Generation IV reactors be different from today's reactors? Does this mean that present day reactors are unsafe, unsustainable and dirty? (Has Gen-IV got anything to do with nuclear fusion? But why should we back both generation IV and fusion ... if fusion is successful, won't this make Gen-IV obsolete?)
- Will Generation IV reactors help us to achieve our ambitious CO2 reduction targets, improve security of energy supply/competitiveness?
- So, Generation IV means widespread use of fast breeder reactors and therefore reprocessing...but won't this mean we need more large reprocessing plants like at La Hague and Sellafield, leading to more low-level radioactive waste and increased radioactive effluent discharges to the environment? These new reprocessing plants will enable all the minor actinides produced in the reactor to be recycled back into fresh fuel...this might be good to reduce the proliferation risk, but won't this lead to increased radiation exposure of workers in both the reprocessing and fuel fabrication plants? But surely we cannot recycle everything...exactly how much highly toxic radioactive waste will be produced each year by a typical Generation IV reactor, and how does this compare with current reactors? Won't this mean we'll still need to find a solution to long-term management of such waste, e.g. construction and operation of geological disposal facilities?
- The availability and extent of uranium reserves and the associated cost is a controversial issue; is there any impact on the deployment strategy of Generation-IV systems?
- What about the use of Thorium in the nuclear fuel cycle?
- What is the GIF?
- What is the GNEP?
- What are the differences and/or links between the GIF and GNEP?
- What is INPRO?
- Is there collaboration between the GIF and INPRO?
- There must be many possible advanced reactor designs ... Why is the GIF looking at only six of these?
- On what basis were the six systems chosen for the R&D phase?
- Will the six systems be safer than existing systems? Will the six systems be cheaper to run than existing systems?
- What inherent aspects of the systems under development will guard against nuclear proliferation?
- Will all six systems be ready for industrial deployment by 2030? When will the first prototype reactor be built? By whom? Is the intention to exploit all six of these Generation IV reactor types on a commercial basis?
- What is the difference between the fast neutron reactors developed in the past, and those to be developed within the Gen-IV framework?
- Does the GIF have any legal basis? How does it operate?
- How is the GIF financed?
- Shouldn't it be the nuclear industry's job to develop and commercialize new reactors?
- How does a country become a member of the GIF? Why are some countries listed as being "non-active" members of the GIF? Is there any concern that these "non-active" members will withdraw completely from the GIF?
- Where are the GIF offices? What is the NEA's role in the GIF?
Watch an Introduction to Generation IV Nuclear Energy Systems and the International Forum (requires Adobe Flash Player, or download in pdf format, 1.2 mb)
Download an Overview of the GIF (pdf, 161 kb)
GIF and Generation-IV
http://www.gen-4.org/PDFs/GIF_Overview.pdf
Motivations and objectives
The Generation IV International Forum, or GIF, was chartered in July 2001 to lead the
collaborative efforts of the world's leading nuclear technology nations to develop next
generation nuclear energy systems to meet the world's future energy needs.
Taking into account the expected increase in energy demand worldwide and the growing
awareness about global warming, climate change issues and sustainable development, nuclear
energy will be needed to meet future global energy demand.
Nuclear power plant technology has evolved as distinct design generations:
First Generation: prototypes, and first realisations (~1950-1970)
Second Generation: current operating plants (~1970-2030)
Third generation: deployable improvements to current reactors (~2000 and on).
Fourth generation: advanced and new reactor systems ( 2030 and beyond)
Eight technology goals have been defined for Generation IV systems in four broad areas:
sustainability, economics, safety and reliability, and proliferation resistance and physical
protection (http://gifdev/PDFs/GenIVRoadmap.pdf). These ambitious goals are shared by a large
number of countries as they aim at responding to economic, environmental and social
requirements of the 21st century. They establish a framework and identify concrete targets for
focusing GIF R&D efforts
Goals for Generation IV Nuclear Energy Systems
Sustainability-1 Generation IV nuclear energy systems will provide sustainable energy generation
that meets clean air objectives and provides long-term availability of systems and effective fuel utilization
for worldwide energy production.
Sustainability-2 Generation IV nuclear energy systems will minimize and manage their nuclear
waste and notably reduce the long-term stewardship burden, thereby improving protection for the public
health and the environment.
Economics-1 Generation IV nuclear energy systems will have a clear life-cycle cost advantage
over other energy sources.
Economics-2 Generation IV nuclear energy systems will have a level of financial risk
comparable to other energy projects.
Safety and Reliability-1 Generation IV nuclear energy systems operations will excel in safety and
reliability.
Safety and Reliability-2 Generation IV nuclear systems will have a very low likelihood and degree of
reactor core damage.
Safety and Reliability-3 Generation IV nuclear energy systems will eliminate the need for offsite
emergency response.
Proliferation resistance and Physical Protection Generation IV nuclear energy systems will
increase the assurance that they are very unattractive and the least desirable route for diversion or theft of
weapons-usable materials, and provide increased physical protection against acts of terrorism.
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