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GEN IV Reactors

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Generation IV Nuclear Reactors

(July 2006)

  • An international task force is developing six nuclear reactor technologies for deployment between 2010 and 2030.
  • 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.

After some two years' deliberation, the Generation IV International Forum (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 are 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 will be the subject of further development internationally.

The GIF was initiated in 2000 and formally chartered in mid 2001. It is an international collective representing governments of countries where nuclear energy is significant now and also seen as vital for the future. They are committed to joint development of the next generation of nuclear technology. Led by the USA, Argentina, Brazil, Canada, France, Japan, South Korea, South Africa, Switzerland, and the UK are members of the GIF, along with the EU. Russia and China were admitted in 2006.

In addition to selecting these six concepts for deployment between 2010 and 2030, the GIF recognized 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 maximize the resource base and minimize high-level wastes to be sent to a repository. Three of the six are fast reactors 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 may mean that they can be in commercial operation well before 2030.

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.

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 utilize thorium as a nuclear fuel. A three-stage program has the first stage well-established, with Pressurized 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.

GIF Reactor technologies:

Gas-cooled fast reactors. Like other helium-cooled reactors which have operated or are under development, these will be high-temperature units - 850°C, suitable for power generation, thermo chemical hydrogen production or other process heat. For electricity, the gas will directly drive a gas turbine (Brayton cycle). Fuels would include depleted uranium and any other fissile or fertile materials. Spent fuel would be reprocessed on site and all the actinides recycled to minimize 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.

Lead-cooled fast reactors. Liquid metal (Pb or Pb-Bi) cooling is by natural convection. Fuel is depleted uranium 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 and this would enable thermochemical hydrogen production.

This corresponds with Russia's BREST fast reactor technology which is lead-cooled and builds on 40 years experience of lead-bismuth cooling 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.

Molten salt reactors. 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. Fission products are removed continuously and the actinides are fully recycled, while plutonium and other actinides can be added along with U-238. 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.

During the 1960s the USA developed the molten salt breeder reactor as the primary back-up option for the conventional fast breeder reactor and a small prototype was operated. 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.

Sodium-cooled fast reactors. This builds on more than 300 reactor-years experienced with fast neutron reactors over five decades and in eight countries. It utilises depleted uranium in the fuel and has a coolant temperature of 550°C enabling electricity generation via a secondary sodium circuit, the primary one being at near atmospheric pressure. Two variants are proposed: a 150-500 MWe type with actinides incorporated into a metal fuel requiring pyrometallurgical processing on site, and a 500-1500 MWe type with conventional MOX fuel reprocessed in conventional facilities elsewhere.

Supercritical water-cooled reactors. This is a very high-pressure water-cooled reactor which operates above the thermodynamic critical point of water 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. 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. Most research on the design has been in Japan.

Very high-temperature gas reactors. These are graphite-moderated, helium-cooled reactors, based on substantial experience. 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 and the PBMR under development in South Africa, with international partners. Outlet temperature of 1000°C 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. Modules of 600 MW thermal are envisaged.

 

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

288

electricity & hydrogen

Lead-cooled fast reactors

fast

Pb-Bi

550-800

low

U-238 +

closed, regional

50-150**
300-400
1200

electricity & hydrogen

Molten salt reactors

epithermal

fluoride salts

700-800

low

UF in salt

closed

1000

electricity & hydrogen

Sodium-cooled fast reactors

fast

sodium

550

low

U-238 & MOX

closed

150-500
500-1500

electricity

Supercritical water-cooled reactors

thermal or fast

water

510-550

very high

UO2

open (thermal)
closed (fast)

1500

electricity

Very high temperature gas reactors

thermal

helium

1000

high

UO2
prism or pebbles

open

250

hydrogen & electricity

The Gas-Cooled Fast Reactor (GFR) system features a fast-neutron-spectrum, helium-cooled reactor and closed fuel cycle.

Like thermal-spectrum, helium-cooled reactors, the high outlet temperature of the helium coolant makes it possible to deliver electricity, hydrogen, or process heat with high efficiency. The reference reactor is a 288-MWe helium-cooled system operating with an outlet temperature of 850 degrees Celsius using a direct Brayton cycle gas turbine for high thermal efficiency.

Several fuel forms are candidates that hold the potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations may be based on prismatic blocks, pin- or plate-based assemblies. The GFR reference has an integrated, on-site spent fuel treatment and refabrication plant.

The GFR uses a direct-cycle helium turbine for electricity generation, or can optionally use its process heat for thermochemical production of hydrogen. Through the combination of a fast spectrum and full recycle of actinides, the GFR minimizes the production of long-lived radioactive waste. The GFR's fast spectrum also makes it possible to use available fissile and fertile materials (including depleted uranium) considerably more efficiently than thermal spectrum gas reactors with once-through fuel cycles.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

The Lead-Cooled Fast Reactor (LFR) system features a fast-spectrum lead or lead/bismuth eutectic liquid-metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.

The system has a full actinide recycle fuel cycle with central or regional fuel cycle facilities. Options include a range of plant ratings, including a battery of 50-150 MWe that features a very long refueling interval, a modular system rated at 300-400 MWe, and a large monolithic plant option at 1200 MWe. The term battery refers to the long-life, factory fabricated core, not to any provision for electrochemical energy conversion. The fuel is metal or nitride-based, containing fertile uranium and transuranics. The LFR is cooled by natural convection with a reactor outlet coolant temperature of 550 degrees C, possibly ranging up to 800 degrees C with advanced materials. The higher temperature enables the production of hydrogen by thermochemical processes.

The LFR battery is a small factory-built turnkey plant operating on a closed fuel cycle with very long refueling interval (15 to 20 years) cassette core or replaceable reactor module. Its features are designed to meet market opportunities for electricity production on small grids, and for developing countries that may not wish to deploy an indigenous fuel cycle infrastructure to support their nuclear energy systems. The battery system is designed for distributed generation of electricity and other energy products, including hydrogen and potable water.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

The Molten Salt Reactor (MSR) system produces fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full actinide recycle fuel cycle.

In the MSR system, the fuel is a circulating liquid mixture of sodium, zirconium, and uranium fluorides. The molten salt fuel flows through graphite core channels, producing an epithermal spectrum. The heat generated in the molten salt is transferred to a secondary coolant system through an intermediate heat exchanger, and then through a tertiary heat exchanger to the power conversion system. The reference plant has a power level of 1,000 MWe. The system has a coolant outlet temperature of 700 degrees Celsius, possibly ranging up to 800 degrees Celsius, affording improved thermal efficiency.

The closed fuel cycle can be tailored to the efficient burn up of plutonium and minor actinides. The MSR's liquid fuel allows addition of actinides such as plutonium and avoids the need for fuel fabrication. Actinides - and most fission products - form fluorinides in the liquid coolant. Molten fluoride salts have excellent heat transfer characteristics and a very low vapor pressure, which reduce stresses on the vessel and piping.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

The Sodium-Cooled Fast Reactor (SFR) system features a fast-spectrum, sodium-cooled reactor and a closed fuel cycle for efficient management of actinides and conversion of fertile uranium.

The fuel cycle employs a full actinide recycle with two major options: One is an intermediate size (150 to 600 MWe) sodium-cooled reactor with uranium-plutonium-minor-actinide-zirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in facilities integrated with the reactor. The second is a medium to large (500 to 1,500 MWe) sodium-cooled reactor with mixed uranium-plutonium oxide fuel, supported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. The outlet temperature is approximately 550 degrees Celsius for both.

The SFR is designed for management of high-level wastes and, in particular, management of plutonium and other actinides. Important safety features of the system include a long thermal response time, a large margin to coolant boiling, a primary system that operates near atmospheric pressure, and intermediate sodium system between the radioactive sodium in the primary system and the water and steam in the power plant. With innovations to reduce capital cost, the SFR can serve markets for electricity.

The SFR's fast spectrum also makes it possible to use available fissile and fertile materials (including depleted uranium) considerably more efficiently than thermal spectrum reactors with once-through fuel cycles.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

The Supercritical-Water-Cooled Reactor (SCWR) system is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 degrees Celsius, 22.1 MPa, or 705 degrees Fahrenheit, 3208 psia).

The supercritical water coolant enables a thermal efficiency about one-third higher than current light-water reactors, as well as simplification in the balance of plant. The balance of plant is considerably simplified because the coolant does not change phase in the reactor and is directly coupled to the energy conversion equipment. The reference system is 1,700 MWe with an operating pressure of 25 MPa, and a reactor outlet temperature of 510 degrees Celsius, possibly ranging up to 550 degrees Celsius. The fuel is uranium oxide. Passive safety features are incorporated similar to those of simplified boiling water reactors.

The SCWR system is primarily designed for efficient electricity production, with an option for actinide management based on two options in the core design: the SCWR may have a thermal or fast-spectrum reactor; the second is a closed cycle with a fast-spectrum reactor and full actinide recycle based on advanced aqueous processing at a central location.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

The Very-High-Temperature Reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a thermal neutron spectrum.

The VHTR is designed to be a high-efficiency system, which can supply electricity and process heat to a broad spectrum of high-temperature and energy-intensive processes.

The reference reactor is a 600 MWth core connected to an intermediate heat exchanger to deliver process heat. The reactor core can be a prismatic block core or a pebble-bed core according to the fuel particles assembly. Fuel particles are coated with successive material layers, high temperature resistant, then formed either into fuel compacts embedded in graphite block for the prismatic block-type core reactor, or formed into graphite coated pebbles. The reactor supplies heat with core outlet temperatures up to 1,000 degrees Celsius, which enables such applications as hydrogen production or process heat for the petrochemical industry. As a nuclear heat application, hydrogen can be efficiently produced from only heat and water by using thermochemical iodine-sulfur process, or high temperature electrolysis process or with additional natural gas by applying the steam reformer technology.

Thus, the VHTR offers a high-efficiency electricity production and a broad range of process heat applications, while retaining the desirable safety characteristics in normal as well as off-normal events. Solutions to adequate waste management will be developed. The basic technology for the VHTR has been well established in former High Temperature Gas Reactors plants, such as the US Fort Saint Vrain and Peach Bottom prototypes, and the German AVR and THTR prototypes. The technology is being advanced through near or medium term projects lead by several plant vendors and national laboratories, such as: PBMR, GT-HTR300C, ANTARES, NHDD, GT-MHR and NGNP in South Africa, Japan, France, South Korea and the United States. Experimental reactors: HTTR (Japan, 30 MWth) and HTR-10 (China, 10 MWth) support the advanced concept development, and the cogeneration of electricity and nuclear heat application.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

 

Technology Roadmap for Generation IV Nuclear Energy Systems

The technology roadmap defines and plans the necessary research and development (R&D) to support the next generation of innovative nuclear energy systems known as Generation IV. The roadmap has been an international effort of ten countries, including Argentina, Brazil, Canada, France, Japan, Korea, South Africa, Switzerland, the United Kingdom, and the United States, the International Atomic Energy Agency, and the OECD Nuclear Energy Agency.

Beginning in 2001, over 100 experts from these countries and international organizations began work on defining the goals for new systems, identifying many promising concepts, and evaluating them, and defining the R&D needed for the most promising systems. By the end of 2002, the work resulted in a description of the six most promising systems and their associated R&D needs. The six systems feature increased safety, improved economics for electricity production and new products such as hydrogen for transportation applications, reduced nuclear wastes for disposal, and increased proliferation resistance.

 

 

 

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