The Generation IV International Forum (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.
After some two years’ deliberation, 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.
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.
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.
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 harmonized. NRC has published its draft design requirements.
The Generation IV Systems selected in 2002 are:
- Very-High-Temperature Reactor (VHTR): a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle.
The Very-High-Temperature Reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle. It supplies heat with high core outlet temperatures which enables applications such as hydrogen production or process heat for the petrochemical industry or others.
The VHTR system is designed to be a high-efficiency system that can supply process heat to a broad spectrum of high-temperature and energy-intensive, non-electric processes. The system may incorporate electricity generating equipment to meet cogeneration needs. The system also has the flexibility to adopt uranium/plutonium fuel cycles and offer enhanced waste minimization. Thus, the VHTR offers a broad range of process heat applications and an option for high-efficiency electricity production, while retaining the desirable safety characteristics offered by modular high-temperature gas-cooled reactors.
The Next Generation Nuclear Plant (NGNP) prototype concept employs the VHTR concept, and is based on what is judged to be the lowest risk technology development that will achieve the needed commercial functional requirements to provide an economically competitive nuclear heat source and hydrogen production capability. The reference concept includes a helium-cooled, graphite moderated, thermal neutron spectrum reactor. The reactor outlet temperature will be in the range of 850 to 950ÂºC. The reactor core technology will either be a prismatic block or pebble bed concept. The NGNP will produce both electricity and hydrogen using an indirect cycle with an intermediate heat exchanger to transfer the heat to either a hydrogen-production demonstration facility or a gas turbine.
The objectives of the NGNP are to (1) demonstrate safe and economical nuclear-assisted production of hydrogen and electricity, (2) demonstrate the basis for commercialization of the nuclear system, the hydrogen production facility, and the power conversion concept, and to (3) establish the basis for Nuclear Regulatory Commission licensing of the commercial version of NGNP.
- Supercritical-Water-Cooled Reactor (SCWR): a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water.
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 (374Â°C, 22 MPa, or 705Â°F, 3208 psia)
SCWRs are built upon two proven technologies: Light Water Reactors (LWRs), which are the most commonly deployed power-generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world.
SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current LWRs) and considerable plant simplification. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for recirculation and jet pumps, pressurizers, steam generators, and steam separators and dryers in current LWRs is eliminated.
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 first option is an SCWR with a thermal or fast-spectrum reactor; the second option is a closed cycle with a fast-spectrum reactor and full actinide recycle based on advanced aqueous processing at a central location.
- Gas-Cooled Fast Reactor (GFR): features a fast-neutron-spectrum, helium-cooled reactor and closed fuel cycle.
The Gas-Cooled Fast Reactor (GFR) system features a fast-spectrum, helium-cooled reactor and closed fuel cycle. The main characteristics of the GFR are: a self-generating core (i.e., conversion ratio = 1) with a fast neutron spectrum, robust refractory fuel, high operating temperature, direct energy conversion with a gas turbine, and full actinide recycling (possibly with an integrated, on-site fuel reprocessing facility).
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 is primarily envisioned for missions in electricity production and actinide management, although it may be able to support hydrogen production as well. The GFR design will utilize a direct-cycle, helium turbine for electricity and process heat for 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.
- Lead-Cooled Fast Reactor (LFR): features a fast-spectrum lead of lead/bismuth eutectic liquid metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.
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 lead (Pb) coolant exhibits very low parasitic absorption of fast neutrons, and this enables the sustainability and fuel cycle benefits traditionally associated with liquid metal-cooled fast spectrum reactors. Pb does not react readily with air, water/steam, or carbon dioxide, eliminating concerns about vigorous exothermic reactions. It has a high boiling temperature (1,740 C), so the need to operate under high pressure and the prospect of boiling or flashing in case of pressure reduction are eliminated.
The LFR is mainly envisioned for electricity and hydrogen production and actinide management. Options for the LFR include a range of plant ratings and sizes from small modular systems to multi-hundred megawatt sized plants. Two key technical aspects of the LFR that offer the prospect for achieving non-proliferation, sustainability, safety and reliability, and economics goals are the use of Pb coolant and a long-life, cartridge-core architecture in a small, modular system intended for deployment with small grids or remote locations. Some technologies for the LFR have already been successfully demonstrated internationally.
- Sodium-Cooled Fast Reactor (SFR): features a fast-spectrum, sodium-cooled reactor and closed fuel cycle for efficient management of actinides and conversion of fertile uranium.
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.
A range of plant size options is available for the SFR, from small modular systems of 50 MWe to large monolithic reactors of approximately 1,500 MWe. The primary coolant system in a SFR can either be arranged in a pool layout – where all primary system components are housed in a single vessel or in a compact loop. For both options, there is a relatively large thermal inertia of the primary coolant. A large margin to coolant boiling is achieved by design and is an important safety feature of these systems. Another major safety feature is that the primary system operates at essentially atmospheric pressure. A secondary sodium system acts as a buffer between the radioactive sodium in the primary system and the energy conversion system in the power plant.
The two main fuel options for the SFR are: (1) mixed uranium-plutonium oxide (MOX) or (2) mixed uranium-plutonium-zirconium metal alloy (metal). The international experience with MOX fuel is more extensive, while the metal fuel offers advantages in safety performance. Other advanced options being considered are nitride, carbide, or dispersion fuels.
The primary mission for the SFR is actinide management for improved waste disposal and uranium resource utilization. With innovations to reduce capital cost, the mission can extend to electricity production and/or heat supply alternatives (hydrogen production, desalination, etc.).
- Molten Salt Reactor (MSR): produces fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full actinide recycle fuel cycle.
Molten Salt Reactors (MSRs) are liquid-fueled reactors that can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels.
Electricity production and waste burndown are envisioned as the primary missions for the MSR. Fissile, fertile, and fission isotopes are dissolved in a high-temperature molten fluoride salt with a very high boiling point (1,400 C) that is both the reactor fuel and the coolant. The near-atmospheric-pressure molten fuel salt flows through the reactor core. The traditional MSR designs have a graphite core that results in a thermal to epithermal neutron spectrum.
In the core, fission occurs within the flowing fuel salt that is heated to ~700ÂºC, which then flows into a primary heat exchanger where the heat is transferred to a secondary molten salt coolant. The fuel salt then flows back to the reactor core. The clean salt in the secondary heat transport system transfers the heat from the primary heat exchanger to a high-temperature Brayton cycle that converts the heat to electricity. The Brayton cycle (with or without a steam bottoming cycle) may use either nitrogen or helium as a working gas.