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61-425: No precise definition of a Generation IV reactor exists. The term refers to nuclear reactor technologies under development as of approximately 2000, and whose designs were intended to represent 'the future shape of nuclear energy', at least at that time. The six designs selected were: the gas-cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR),

122-450: A neutron moderator to slow neutrons until they approach the average kinetic energy of the surrounding particles, that is, to reduce the speed of the neutrons to low-velocity, thermal neutrons. Neutrons are uncharged, this allows them to penetrate deep in the target and close to the nuclei, thus scattering neutrons by nuclear forces, some nuclides are scattered large. The nuclear cross section of uranium-235 for slow thermal neutrons

183-587: A U.S. research laboratory put it, "fabrication, construction, operation, and maintenance of new reactors will face a steep learning curve: advanced technologies will have a heightened risk of accidents and mistakes. The technology may be proven, but people are not". Sodium fast reactor A sodium-cooled fast reactor is a fast neutron reactor cooled by liquid sodium . The initials SFR in particular refer to two Generation IV reactor proposals, one based on existing liquid metal cooled reactor (LMFR) technology using mixed oxide fuel (MOX), and one based on

244-410: A core with a graphite moderator . The fuel may be dispersed in a graphite matrix. These designs are more accurately termed an epithermal reactor than a thermal reactor due to the higher average speed of the neutrons that cause the fission events. MCSFR does away with the graphite moderator. They achieve criticality using a sufficient volume of salt and fissile material. They can consume much more of

305-745: A demonstration HTR-PM 200-MW high temperature pebble bed reactor as a successor to its HTR-10 . A molten salt reactor (MSR) is a type of reactor where the primary coolant or the fuel itself is a molten salt mixture. It operates at high temperature and low pressure. Molten salt can be used for thermal, epithermal and fast reactors. Since 2005 the focus has been on fast spectrum MSRs (MSFR). Other designs include integral molten salt reactors (e.g. IMSR) and molten chloride salt fast reactors (MCSFR). Early thermal spectrum concepts and many current ones rely on uranium tetrafluoride (UF 4 ) or thorium tetrafluoride (ThF 4 ), dissolved in molten fluoride salt. The fluid reaches criticality by flowing into

366-537: A design similar to Areva 's prismatic block Antares reactor to be deployed as a prototype by 2021. In January 2016, X-energy was provided a five-year grant of up to $ 40 million by the United States Department of Energy to advance their reactor development. The Xe-100 is a PBMR that would generate 80 MWe , or 320 MWe in a 'four-pack'. Since 2021, the Chinese government is operating

427-456: A fuel cycle based on pyrometallurgical reprocessing in facilities integrated with the reactor. The second is a medium to large (500–1,500 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 multiple reactors. The outlet temperature is approximately 510–550 degrees C for both. Liquid metallic sodium may be used to carry heat from

488-541: A half-life of only 15 hours. Another problem is leaks. Sodium at high temperatures ignites in contact with oxygen. Such sodium fires can be extinguished by powder, or by replacing the air with nitrogen . A Russian breeder reactor, the BN-600, reported 27 sodium leaks in a 17-year period, 14 of which led to sodium fires. No fission products have a half-life in the range of 100 a–210 ka ... ... nor beyond 15.7 Ma The operating temperature must not exceed

549-655: A large margin to coolant boiling, a primary cooling system that operates near atmospheric pressure, and an intermediate sodium system between the radioactive sodium in the primary system and the water and steam in the power plant. Innovations can reduce capital cost, such as modular designs, removing a primary loop, integrating the pump and intermediate heat exchanger, and better materials. The SFR's fast spectrum 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. In 2020 Natrium received an $ 80M grant from

610-480: A large monolithic plant at 1,200 MW e . The fuel is metal or nitride-based containing fertile uranium and transuranics . The reactor is cooled by natural convection with a reactor outlet coolant temperature of 550-800 °C. The higher temperature enables the production of hydrogen by thermochemical processes . The European Sustainable Nuclear Industrial Initiative is funding a 100 MW t LFR, an accelerator-driven sub-critical reactor called MYRRHA . It

671-465: A manner that will provide a competitively priced and reliable supply of energy ... while satisfactorily addressing nuclear safety, waste, proliferation and public perception concerns." It coordinates the development of GEN IV technologies. It has been instrumental in coordinating research into the six types of Generation IV reactors, and in defining the scope and meaning of the term itself. As of 2021, active members include: Australia , Canada , China ,

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732-464: A thermal reactor. It uses supercritical water as the working fluid. SCWRs are basically light water reactors (LWR) operating at higher pressure and temperatures with a direct, once-through heat exchange cycle. As commonly envisioned, it would operate on a direct cycle, much like a boiling water reactor (BWR). Since it uses supercritical water (not to be confused with critical mass ) as the working fluid, it would have only one water phase. This makes

793-615: Is a pebble-bed type high-temperature gas-cooled reactor . It was connected to the grid in December 2023, making it the world's first Gen IV reactor to enter commercial operation. In 2024, it was reported that China would also build the world’s first thorium molten salt nuclear power station, scheduled to be operational by 2029. The Generation IV International Forum (GIF) is an international organization with its stated goal being "the development of concepts for one or more Generation IV systems that can be licensed, constructed, and operated in

854-461: Is about 1000 barns , while for fast neutrons it is in the order of 1 barn. Therefore, thermal neutrons are more likely to cause uranium-235 to nuclear fission than to be captured by uranium-238 . If at least one neutron from the U-235 fission strikes another nucleus and causes it to fission, then the chain reaction will continue. If the reaction will sustain itself, it is said to be critical , and

915-491: Is being built at a cost of INR 5,677 crores (~US$ 900 million). After numerous delays, the government reported in March 2020 that the reactor might be operational in December 2021. The PFBR was to be followed by six more Commercial Fast Breeder Reactors (CFBRs) of 600 MW e each. The Gen IV SFR is a project that builds on the oxide fueled fast breeder reactor and the metal fueled integral fast reactor . Its goals are to increase

976-440: Is below the consumption rate, thus reducing the nuclear storage problem , without the nuclear proliferation concerns and other technical issues associated with a fast reactor . The supercritical water reactor (SCWR) is a reduced moderation water reactor concept. Because the average speed of the fission-causing neutrons within the fuel is faster than thermal neutrons , it is more accurately termed an epithermal reactor than

1037-472: Is cooled by liquid sodium and fueled by a metallic alloy of uranium and plutonium or spent nuclear fuel , the nuclear waste of light water reactors . The SFR fuel is contained in steel cladding. Liquid sodium fills the space between the clad elements that make up the fuel assembly. One of the design challenges is the risks of handling sodium, which reacts explosively if it comes into contact with water. The use of liquid metal instead of water as coolant allows

1098-423: Is limited by the production of plutonium from uranium. One work-around is to have an inert matrix, using, e.g., magnesium oxide . Magnesium oxide has an order of magnitude lower probability of interacting with neutrons (thermal and fast) than elements such as iron. High-level wastes and, in particular, management of plutonium and other actinides must be handled. Safety features include a long thermal response time,

1159-495: Is not possible in thermal reactor. In contrast to thermal-neutron reactors, integral fast reactors (IFRs) operate using fast neutrons and are designed for increased fuel efficiency. These reactors are capable of recycling nuclear waste and breeding new fuel, which enhances sustainability. Additionally, IFRs incorporate passive safety features that allow them to safely shut down without external power or human intervention Most nuclear power plant reactors are thermal reactors and use

1220-515: Is that metal atoms are weak neutron moderators. Water is a much stronger neutron moderator because the hydrogen atoms found in water are much lighter than metal atoms, and therefore neutrons lose more energy in collisions with hydrogen atoms. This makes it difficult to use water as a coolant for a fast reactor because the water tends to slow (moderate) the fast neutrons into thermal neutrons (although concepts for reduced moderation water reactors exist). Another advantage of liquid sodium coolant

1281-403: Is that sodium melts at 371K (98°C) and boils / vaporizes at 1156K (883°C), a difference of 785K (785°C) between solid / frozen and gas / vapor states. By comparison, the liquid temperature range of water (between ice and gas) is just 100K at normal, sea-level atmospheric pressure conditions. Despite sodium's low specific heat (as compared to water), this enables the absorption of significant heat in

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1342-501: Is the VVER -1700/393 (VVER-SCWR or VVER-SKD) – a Russian SCWR with double-inlet-core and a breeding ratio of 0.95. A fast reactor directly uses fission neutrons without moderation. Fast reactors can be configured to "burn", or fission, all actinides , and given enough time, therefore drastically reduce the actinides fraction in spent nuclear fuel produced by the present world fleet of thermal neutron light water reactors , thus closing

1403-664: Is to be built in Belgium with construction expected by 2036. A reduced-power model called Guinevere was started up at Mol in March 2009 and became operational in 2012. Two other lead-cooled fast reactors under development are the SVBR-100, a modular 100 MW e lead-bismuth cooled fast neutron reactor concept designed by OKB Gidropress in Russia and the BREST-OD-300 (Lead-cooled fast reactor) 300 MW e , to be developed after

1464-726: The European Atomic Energy Community (Euratom), France , Japan , Russia , South Africa , South Korea , Switzerland , the United Kingdom and the United States . Non-active members include Argentina and Brazil . The Forum was initiated in January 2000 by the Office of Nuclear Energy of the U.S. Department of Energy ’s (DOE) "as a co-operative international endeavour seeking to develop

1525-723: The Fast Breeder Test Reactor (FBTR) reached criticality in October 1985. In September 2002, fuel burn up efficiency in the FBTR for the first time reached the 100,000 megawatt-days per metric ton uranium (MWd/MTU) mark. This is considered an important milestone in Indian breeder reactor technology. Using that experience, the Prototype Fast Breeder Reactor , a 500 MWe Sodium cooled fast reactor

1586-614: The Natrium appellation in Kemmerer, Wyoming . Aside from the Russian experience, Japan, India, China, France and the USA are investing in the technology. The nuclear fuel cycle employs a full actinide recycle with two major options: One is an intermediate-size (150–600 MWe) sodium-cooled reactor with uranium - plutonium -minor-actinide- zirconium metal alloy fuel, supported by

1647-704: The US Department of Energy for development of its SFR. The program plans to use High-Assay, Low Enriched Uranium fuel containing 5-20% uranium. The reactor was expected to be sited underground and have gravity-inserted control rods. Because it operates at atmospheric pressure, a large containment shield is not necessary. Because of its large heat storage capacity, it was expected to be able to produce surge power of 500 MWe for 5+ hours, beyond its continuous power of 345 MWe. Sodium-cooled reactors have included: Most of these were experimental plants that are no longer operational. On November 30, 2019, CTV reported that

1708-425: The loop type Prototype Fast Breeder Reactor Monju at Tsuruga, Japan. Using lead or molten salt coolants mitigates this problem as they are less reactive and have a high freezing temperature and ambient pressure. Lead has much higher viscosity, much higher density, lower heat capacity, and more radioactive neutron activation products than sodium. Multiple proof of concept Gen IV designs have been built. For example,

1769-696: The 1980s. The two largest experimental sodium cooled fast reactors are in Russia, the BN-600 and the BN-800 (880 MWe gross). These NPPs are being used to provide operating experience and technological solutions that will be applied to the construction of the BN-1200 ( OKBM Afrikantov first Gen IV reactor). The largest ever operated was the French Superphenix reactor at over 1200 MW e , successfully operating before decommissioning in 1996. In India,

1830-498: The 20 MW e EBR II operated for over thirty years at Idaho National Laboratory, but was shut down in 1994. GE Hitachi's PRISM reactor is a modernized and commercial implementation of the Integral Fast Reactor (IFR), developed by Argonne National Laboratory between 1984 and 1994. The primary purpose of PRISM is burning up spent nuclear fuel from other reactors, rather than breeding new fuel. The design reduces

1891-482: The Canadian provinces of New Brunswick , Ontario and Saskatchewan planned an announcement about a joint plan to cooperate on small sodium fast modular nuclear reactors from New Brunswick-based ARC Nuclear Canada. Thermal reactor A thermal-neutron reactor is a nuclear reactor that uses slow or thermal neutrons . ("Thermal" does not mean hot in an absolute sense, but means in thermal equilibrium with

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1952-457: The MSR is the possibility of a thermal spectrum nuclear waste-burner . Conventionally only fast spectrum reactors have been considered viable for utilization or reduction of the spent nuclear fuel . Thermal waste-burning was achieved by replacing a fraction of the uranium in the spent nuclear fuel with thorium . The net production rate of transuranic elements (e.g. plutonium and americium )

2013-588: The SVBR-100, it will dispense with the fertile blanket around the core and will supersede the sodium cooled BN-600 reactor design, to purportedly give enhanced proliferation resistance. Preparatory construction work commenced in May 2020. The GEN IV Forum reframes the reactor safety paradigm, from accepting that nuclear accidents can occur and should be mastered, to eliminating the physical possibility of an accident. Active and passive safety systems would be at least as effective as those of Generation III systems and render

2074-524: The commercialisation phases are set. According to the GIF in 2013, "It will take at least two or three decades before the deployment of commercial Gen IV systems." Many reactor types were considered initially; the list was then refined to focus on the most promising technologies. Three systems are nominally thermal reactors and three are fast reactors . The very high temperature reactor (VHTR) potentially can provide high quality process heat. Fast reactors offer

2135-458: The coolant (the Phénix reactor outlet temperature was 833K (560°C)) permit a higher thermodynamic efficiency than in water cooled reactors. The electrically conductive molten sodium can be moved by electromagnetic pumps . The fact that the sodium is not pressurized implies that a much thinner reactor vessel can be used (e.g. 2 cm thick). Combined with the much higher temperatures achieved in

2196-423: The core. Sodium has only one stable isotope, sodium-23 , which is a weak neutron absorber. When it does absorb a neutron it produces sodium-24 , which has a half-life of 15 hours and decays to stable isotope magnesium-24 . The two main design approaches to sodium-cooled reactors are pool type and loop type. In the pool type, the primary coolant is contained in the main reactor vessel, which therefore includes

2257-402: The efficiency of uranium usage by breeding plutonium and eliminating transuranic isotopes. The reactor design uses an unmoderated core running on fast neutrons , designed to allow any transuranic isotope to be consumed (and in some cases used as fuel). SFR fuel expands when the reactor overheats, automatically slowing down the chain reaction, making it passively safe. One SFR reactor concept

2318-516: The efficient production of hydrogen and the synthesis of carbon-neutral fuels . The majority of reactors in operation around the world are considered second generation and third generation reactor systems, as the majority of the first generation systems have been retired. China was the first country to operate a demonstration generation-IV reactor, the HTR-PM in Shidaowan, Shandong , which

2379-440: The fuel and leave only short-lived waste. Most MSR designs are derived from the 1960s Molten-Salt Reactor Experiment (MSRE). Variants include the conceptual Dual fluid reactor that uses lead as a cooling medium with molten salt fuel, commonly a metal chloride, e.g. plutonium(III) chloride , to aid in greater closed-fuel cycle capabilities. Other notable approaches include the stable salt reactor (SSR) concept, which encases

2440-854: The fuel cycle. Alternatively, if configured differently, they can breed more actinide fuel than they consume. The gas-cooled fast reactor (GFR) features a fast-neutron spectrum and closed fuel cycle. The reactor is helium -cooled. Its outlet temperature is 850 °C. It moves the very-high-temperature reactor (VHTR) to a more sustainable fuel cycle. It uses a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are under consideration: composite ceramic fuel, advanced fuel particles, or ceramic-clad actinide compounds. Core configurations involve pin- or plate-based fuel assemblies or prismatic blocks. The European Sustainable Nuclear Industrial Initiative provided funding for three Generation IV reactor systems: Sodium-cooled fast reactors (SCFRs) have been operated in multiple countries since

2501-431: The fuel's boiling temperature. Fuel-to-cladding chemical interaction (FCCI) has to be accommodated. FCCI is eutectic melting between the fuel and the cladding; uranium, plutonium, and lanthanum (a fission product ) inter-diffuse with the iron of the cladding. The alloy that forms has a low eutectic melting temperature. FCCI causes the cladding to reduce in strength and even rupture. The amount of transuranic transmutation

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2562-549: The fuel. The very-high-temperature reactor (VHTR) uses a graphite-moderated core with a once-through uranium fuel cycle, using helium or molten salt. This reactor design envisions an outlet temperature of 1,000°C. The reactor core can be either a prismatic-block or a pebble bed reactor design. The high temperatures enable applications such as process heat or hydrogen production via the thermochemical sulfur-iodine cycle process. In 2012, as part of its next generation nuclear plant competition, Idaho National Laboratory approved

2623-422: The half lives of the fissionable elements present in spent nuclear fuel while generating electricity largely as a byproduct. The lead-cooled fast reactor (LFR) features a fast-neutron-spectrum lead or lead / bismuth eutectic ( LBE ) coolant with a closed fuel cycle . Proposals include a small 50 to 150 MW e that features a long refueling interval, a modular system rated at 300 to 400 MW e , and

2684-492: The heat exchange method more similar to a pressurized water reactor ( PWR ). It could operate at much higher temperatures than both current PWRs and BWRs. Supercritical water-cooled reactors (SCWRs) offer high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current LWRs) and considerable simplification. The mission of the SCWR is generation of low-cost electricity . It is built upon two proven technologies, LWRs,

2745-457: The liquid phase, while maintaining large safety margins. Moreover, the high thermal conductivity of sodium effectively creates a reservoir of heat capacity that provides thermal inertia against overheating. Sodium need not be pressurized since its boiling point is much higher than the reactor's operating temperature , and sodium does not corrode steel reactor parts, and in fact, protects metals from corrosion. The high temperatures reached by

2806-489: The medium it is interacting with, the reactor's fuel, moderator and structure, which is much lower energy than the fast neutrons initially produced by fission.) A fast-neutron reactor , on the other hand, operates using high-energy neutrons that are not slowed by a moderator. These reactors can efficiently use a broader range of fuels, including plutonium and other heavy atoms, and have the capability to breed more fissile material, such as uranium-238 into plutonium-239, which

2867-485: The metal-fueled integral fast reactor . Several sodium-cooled fast reactors have been built and some are in current operation, particularly in Russia. Others are in planning or under construction. For example, in 2022, in the US, TerraPower (using its Traveling Wave technology ) is planning to build its own reactors along with molten salt energy storage in partnership with GEHitachi's PRISM integral fast reactor design, under

2928-453: The molten salt in the well-established fuel rods of conventional reactors. This latter design was found to be the most competitive by consultancy firm Energy Process Development in 2015. Another design under development is TerraPower 's molten chloride fast reactor. This concept mixes the liquid natural uranium and molten chloride coolant in the reactor core, reaching very high temperatures at atmospheric pressure. Another notable feature of

2989-499: The most commonly deployed power generating reactors, and superheated fossil fuel fired boilers , also in wide use. 32 organizations in 13 countries are investigating the concept. SCWRs share the steam explosion and radioactive steam release hazards of BWRs and LWRs as well as the need for extremely expensive heavy duty pressure vessels, pipes, valves, and pumps. These shared problems are inherently more severe for SCWRs due to their higher temperatures. One SCWR design under development

3050-449: The most severe accidents physically impossible. Relative to Gen II-III, advantages of Gen IV reactors include: A specific risk of the SFR is related to using metallic sodium as a coolant. In case of a breach, sodium explosively reacts with water. Argon is used to prevent sodium oxidation. Argon can displace oxygen in the air and can pose hypoxia concerns for workers. This was a factor at

3111-531: The next decade was published in January 2014. In May 2019, Terrestrial Energy , the Canadian developer of a molten salt reactor, became the first private company to join GIF. At the Forum's October 2021 meeting, the Forum members agreed to create a task force on non-electric applications of nuclear heat, including district and industrial heat applications, desalination and large-scale hydrogen production. The GIF Forum has introduced development timelines for each of

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3172-451: The possibility of burning actinides to further reduce waste and can breed more fuel than they consume. These systems offer significant advances in sustainability, safety and reliability, economics, proliferation resistance, and physical protection. A thermal reactor is a nuclear reactor that uses slow or thermal neutrons . A neutron moderator is used to slow the neutrons emitted by fission to make them more likely to be captured by

3233-495: The reaction. A disadvantage of sodium is its chemical reactivity, which requires special precautions to prevent and suppress fires. If sodium comes into contact with water it reacts to produce sodium hydroxide and hydrogen, and the hydrogen burns in contact with air. This was the case at the Monju Nuclear Power Plant in a 1995 accident. In addition, neutron capture causes it to become radioactive; albeit with

3294-442: The reactor core and a heat exchanger . The US EBR-2 , French Phénix and others used this approach, and it is used by India's Prototype Fast Breeder Reactor and China's CFR-600 . In the loop type, the heat exchangers are outside the reactor tank. The French Rapsodie , British Prototype Fast Reactor and others used this approach. All fast reactors have several advantages over the current fleet of water based reactors in that

3355-419: The reactor, this means that the reactor in shutdown mode can be passively cooled. For example, air ducts can be engineered so that all the decay heat after shutdown is removed by natural convection, and no pumping action is required. Reactors of this type are self-controlling. If the temperature of the core increases, the core will expand slightly, which means that more neutrons will escape the core, slowing down

3416-525: The reactors at Fort St. Vrain Generating Station and HTR-10 are similar to the proposed Gen IV VHTR designs, and the pool type EBR-II , Phénix , BN-600 and BN-800 reactor are similar to the proposed pool type Gen IV SFR designs. Nuclear engineer David Lochbaum cautions, "the problem with new reactors and accidents is twofold: scenarios arise that are impossible to plan for in simulations; and humans make mistakes". As one director of

3477-440: The research necessary to test the feasibility and performance of fourth generation nuclear systems, and to make them available for industrial deployment by 2030." It was established in 2001, aiming at availability for industrial deployment by 2030. In November 2013, a brief overview of the reactor designs and activities by each forum member was made available. An update of the technology roadmap which details R&D objectives for

3538-467: The six systems. Research and development is divided into three phases: In 2000, GIF stated, "After the performance phase is complete for each system, at least six years and several US$ billion will be required for detailed design and construction of a demonstration system." In the Roadmap update of 2013, the performance and demonstration phases were considerably shifted to later dates, while no targets for

3599-539: The supercritical-water-cooled reactor (SCWR) and the very high-temperature reactor (VHTR). The sodium fast reactor has received the greatest share of funding that supports demonstration facilities. Moir and Teller consider the molten-salt reactor , a less developed technology, as potentially having the greatest inherent safety of the six models. The very-high-temperature reactor designs operate at much higher temperatures than prior generations. This allows for high temperature electrolysis or for sulfur–iodine cycle for

3660-554: The system to work at atmospheric pressure, reducing the risk of leakage. The European Sustainable Nuclear Industrial Initiative funded three Generation IV reactor systems. Advanced Sodium Technical Reactor for Industrial Demonstration ( ASTRID ) was a sodium-cooled fast reactor, that was cancelled in August 2019. Numerous progenitors of the Gen IV SFR exist. The 400 MW t Fast Flux Test Facility operated for ten years at Hanford;

3721-501: The waste streams are significantly reduced. Crucially, when a reactor runs on fast neutrons, the plutonium isotopes are far more likely to fission upon absorbing a neutron. Thus, fast neutrons have a smaller chance of being captured by the uranium and plutonium, but when they are captured, have a much bigger chance of causing a fission. This means that the inventory of transuranic waste is non existent from fast reactors. The primary advantage of liquid metal coolants, such as liquid sodium,

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