Generation IV reactor

From Wikipedia, the free encyclopedia
  (Redirected from Generation IV reactors)
Jump to: navigation, search
Nuclear Energy Systems Deployable no later than 2030 and offering significant advances in sustainability, safety and reliability, and economics

Generation IV reactors (Gen IV) are a set of mostly theoretical nuclear reactor designs currently being researched. Most of these designs, with the exception of the BN-1200 reactor, are generally not expected to be available for commercial construction before 2030.[1] Most reactors in operation around the world are generally considered second generation reactor systems, with most of the first-generation systems having been retired some time ago while there are only a dozen or so Generation III reactors in operation(2014). Generation V reactors refer to reactors that may be possible but are not yet considered feasible in the short term, and are therefore not receiving as much R&D funding.

Reactor types[edit]

Many reactor types were considered initially; however, the list was downsized to focus on the most promising technologies and those that could most likely meet the goals of the Gen IV initiative.[1] Three systems are nominally thermal reactors and three are fast reactors. The Very High Temperature Reactor (VHTR) is also being researched for potentially providing high quality process heat for hydrogen production. The fast reactors offer the possibility of burning actinides to further reduce waste and of being able to "breed more fuel" than they consume. These systems offer significant advances in sustainability, safety and reliability, economics, proliferation resistance (depending on perspective) and physical protection.

Thermal reactors[edit]

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 the fuel.

Very-high-temperature reactor (VHTR)[edit]

Very-High-Temperature Reactor (VHTR)

The very high temperature reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using helium or molten salt as the coolant. 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 iodine-sulfur process. It would also be passively safe.

The planned construction of the first VHTR, the South African PBMR (pebble bed modular reactor), lost government funding in February, 2010.[2] A pronounced increase of costs and concerns about possible unexpected technical problems had discouraged potential investors and customers.

The Peoples Republic of China began construction of a 200-MWe High Temperature Pebble bed reactor in 2012 as a successor to its HTR-10.[3]

Also in 2012, as part of the Next Generation Nuclear Plant competition, Idaho National Laboratory approved a design similar to Areva's prismatic block Antares reactor as the chosen HTGR to be deployed as a prototype by 2021. It was in competition with General Atomics' Gas turbine modular helium reactor and Westinghouse's Pebble Bed Modular Reactor.[4]

Molten-salt reactor (MSR)[edit]

Molten Salt Reactor (MSR)
Main article: Molten salt reactor

A molten salt reactor[5] is a type of nuclear reactor where the primary coolant, or even the fuel itself is a molten salt mixture. There have been many designs put forward for this type of reactor and a few prototypes built. The early concepts and many current ones rely on nuclear fuel dissolved in the molten fluoride salt as uranium tetrafluoride (UF4) or thorium tetrafluoride (ThF4). The fluid would reach criticality by flowing into a graphite core which would also serve as the moderator. Many current concepts rely on fuel that is dispersed in a graphite matrix with the molten salt providing low pressure, high temperature cooling.

The Gen IV MSR is more accurately termed an epithermal reactor than a thermal reactor due to the average speed of the neutrons that would cause the fission events within its fuel being faster than thermal neutrons.[6]

The principle of a MSR can be used for thermal, epithermal and fast reactors. Since 2005 the focus has moved towards a fast spectrum MSR (MSFR). [7]

Supercritical-water-cooled reactor (SCWR)[edit]

Supercritical-Water-Cooled Reactor (SCWR)

The supercritical water reactor (SCWR)[5] is a reduced moderation water reactor concept that, due to the average speed of the neutrons that would cause the fission events within the fuel being faster than thermal neutrons, it is more accurately termed an epithermal reactor than 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 most commonly envisioned, it would operate on a direct cycle, much like a boiling water reactor (BWR), but since it uses supercritical water (not to be confused with critical mass) as the working fluid, it would have only one water phase present, which makes the supercritical 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) are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current LWRs) and considerable plant simplification.

The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil fuel fired boilers, a large number of which are also in use around the world. The SCWR concept is being investigated by 32 organizations in 13 countries.[citation needed]

A SCWR Design under development is the VVER-1700/393 (VVER-SCWR or VVER-SKD) — a Russian Supercritical-water-cooled reactor with double-inlet-core and a breeding ratio of 0.95.[8]

Fast reactors[edit]

A fast reactor directly uses the fast neutrons emitted by fission, without moderation. Unlike thermal neutron reactors, fast neutron 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 the nuclear fuel cycle. Alternatively, if configured differently, they can also breed more actinide fuel than they consume.

Gas-cooled fast reactor (GFR)[edit]

Gas-Cooled Fast Reactor (GFR)

The gas-cooled fast reactor (GFR)[5] system features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reactor is helium-cooled and with an outlet temperature of 850 °C it is an evolution of the very-high-temperature reactor (VHTR) to a more sustainable fuel cycle. It will use a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their 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 are being considered based on pin- or plate-based fuel assemblies or prismatic blocks.

The European Sustainable Nuclear Industrial Initiative is funding three Generation IV reactor systems, one of which is a gas-cooled fast reactor, called Allegro, 100 MW(t), which will be built in a central or eastern European country with construction expected to begin in 2018.[9] The central European Visegrád Group are committed to pursuing the technology.[10] In 2013 German, British, and French institutes finished a 3 year collaboration study on the follow on industrial scale design, known as GoFastR.[11] They were funded by the EU's 7th FWP framework programme, with the goal of making a sustainable VHTR.[12]

Sodium-cooled fast reactor (SFR)[edit]

Pool design Sodium-Cooled Fast Reactor (SFR)

The SFR[5] is a project that builds on two closely related existing projects, the liquid metal fast breeder reactor and the integral fast reactor.

The goals are to increase the efficiency of uranium usage by breeding plutonium and eliminating the need for transuranic isotopes ever to leave the site. 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). In addition to the benefits of removing the long half-life transuranics from the waste cycle, the SFR fuel expands when the reactor overheats, and the chain reaction automatically slows down. In this manner, it is passively safe.

The SFR reactor concept 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 with liquid sodium filling in the space between the clad elements which make up the fuel assembly. One of the design challenges of an SFR is the risks of handling sodium, which reacts explosively if it comes into contact with water. However, the use of liquid metal instead of water as coolant allows the system to work at atmospheric pressure, reducing the risk of leakage.

The European Sustainable Nuclear Industrial Initiative is funding three Generation IV reactor systems, one of which is a sodium-cooled fast reactor, called ASTRID, Advanced Sodium Technical Reactor for Industrial Demonstration, Areva, CEA and EDF are leading the design with British collaboration.[13][14] Astrid will be rated about 600 MWe and is expected to be built in France, with construction slated to begin in 2017 near to the Phénix reactor.[9]

The PRC's first commercial-scale, 800 MWe, fast neutron reactor, to be situated near Sanming in Fujian province will be a SFR. In 2009 an agreement was signed that would entail the Russian BN-800 reactor design to be sold to the PRC once it is completed, this would be the first time commercial-scale fast neutron reactors have ever been exported.[15] The BN-800 reactor became operational in 2014.

In India, the Prototype Fast Breeder Reactor, a 500MWe Sodium cooled fast reactor is under construction, with a completion year of 2014/2015.

Lead-cooled fast reactor (LFR)[edit]

Lead-Cooled Fast Reactor (LFR)
See also: MYRRHA

The lead-cooled fast reactor[5] features a fast-neutron-spectrum lead or lead/bismuth eutectic (LBE) liquid-metal-cooled reactor with a closed fuel cycle. Options include a range of plant ratings, including a "battery" of 50 to 150 MW of electricity that features a very long refueling interval, a modular system rated at 300 to 400 MW, and a large monolithic plant option at 1,200 MW. (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 °C, possibly ranging up to 800 °C with advanced materials. The higher temperature enables the production of hydrogen by thermochemical processes.

The European Sustainable Nuclear Industrial Initiative is funding three Generation IV reactor systems, one of which is a lead-cooled fast reactor that is also an accelerator-driven sub-critical reactor, called Myrrha, 100 MW(t), which will be built in Belgium with construction expected to begin after 2014 and the industrial scale version, known as Alfred, slated to be constructed sometime after 2017. A reduced-power model of Myrrha called Guinevere was started up at Mol in March 2009.[9] In 2012 the research team reported that Guinevere was operational.[16]

Two other lead-cooled fast reactors under development are the SVBR-100, a modular 100MWe lead-bismuth cooled fast neutron reactor concept designed by OKB Gidropress in Russia and the BREST-OD-300 (Lead-cooled fast reactor) 300 MWe, to be developed after the SVBR-100, and built over 2016-20, 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.[8]

Advantages and disadvantages[edit]

Relative to current nuclear power plant technology, the claimed benefits for 4th generation reactors include:

  • Nuclear waste that remains radioactive for a few centuries instead of millennia [17]
  • 100-300 times more energy yield from the same amount of nuclear fuel [18]
  • Broader range of fuels, and even unencapsulated raw fuels (non-pebble MSR, LFTR).
  • In some reactors, the ability to consume existing nuclear waste in the production of electricity, that is, a Closed nuclear fuel cycle. This strengthens the argument to deem nuclear power as renewable energy.
  • Improved operating safety features, such as (depending on design) avoidance of pressurized operation, automatic passive (unpowered, uncommanded) reactor shutdown, avoidance of water cooling and the associated risks of loss of water (leaks or boiling) and hydrogen generation/explosion and contamination of coolant water.

Nuclear reactors do not emit CO2 during operation, although like all low carbon power sources, the mining and construction phase can result in CO2 emissions, if energy sources which are not carbon neutral (such as fossil fuels), or CO2 emitting cements are used during the construction process. A 2012 Yale University review published in the Journal of Industrial Ecology analyzing CO
life cycle assessment (LCA) emissions from nuclear power determined that:[19]

"The collective LCA literature indicates that life cycle GHG [ greenhouse gas ] emissions from nuclear power are only a fraction of traditional fossil sources and comparable to renewable technologies."

Although the paper primarily dealt with data from Generation II reactors, and did not analyze the CO
emissions by 2050 of the presently under construction Generation III reactors, it did summarize the Life Cycle Assessment findings of in development reactor technologies.

FBRs [ Fast Breeder Reactors ] have been evaluated in the LCA literature. The limited literature that evaluates this potential future technology reports median life cycle GHG emissions... similar to or lower than LWRs[ Gen II light water reactors ] and purports to consume little or no uranium ore.

A specific risk of the sodium-cooled fast reactor is related to using metallic sodium as a coolant. In case of a breach, sodium explosively reacts with water. Fixing breaches may also prove dangerous, as the cheapest noble gas argon is also used to prevent sodium oxidation. Argon, like helium, can displace oxygen in the air and can pose hypoxia concerns, so workers may be exposed to this additional risk. This is a pertinent problem as can be testified by the events at the loop type Prototype Fast Breeder Reactor Monju at Tsuruga, Japan.[20] Using lead or molten salts mitigates this problem by making the coolant less reactive and allowing a high freezing temperature and low pressure in case of a leak.

In many cases, there is already a large amount of experience built up with numerous proof of concept Gen IV designs. For example, 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 and BN-600 reactor are similar to the proposed pool type Gen IV Sodium Cooled Fast reactors being designed.

Generation IV International Forum[edit]

There are currently ten active members of the Generation IV International Forum (GIF): Canada, China, the European Atomic Energy Community (Euratom), France, Japan, Russia, South Africa, South Korea, Switzerland, and the United States. The non-active members are Argentina, Brazil, and the United Kingdom.[21]

The Generation IV International Forum (GIF) was founded in 2001. Switzerland joined in 2002, Euratom in 2003, and China and Russia in 2006. The remaining countries were founding members.[21]

The 36th GIF meeting in Brussels was held in November 2013.[22][23] The Technology Roadmap Update for Generation IV Nuclear Energy Systems was published in January 2014 which details R&D objectives for the next decade.[24] A breakdown of the reactor designs being researched by each forum member has been made available.[25]

See also[edit]


  1. ^ a b Generation IV
  2. ^ South Africa to stop funding Pebble Bed nuclear reactor
  3. ^ Nucnet Report: 'China Begins Construction of First Generation IV HTR-PM Unit', 7 January 2013
  4. ^ "INL approves Antares design". 
  5. ^ a b c d e US DOE Nuclear Energy Research Advisory Committee (2002). "A Technology Roadmap for Generation IV Nuclear Energy Systems". GIF-002-00. 
  6. ^ "Idaho National Laboratory detailing some current efforts at developing Gen. IV reactors". 
  7. ^ H. Boussier, S. Delpech, V. Ghetta et Al. : The Molten Salt Reactor (MSR) in Generation IV: Overview and Perspectives, GIF SYMPOSIUM PROCEEDINGS/2012 ANNUAL REPORT, NEA No. 7141, pp95 [1]
  8. ^ a b "Technology Developments & Plant Efficiency for the Russian Nuclear Power Generation Market Wednesday". March 24, 2010. 
  9. ^ a b c "The European Sustainable Nuclear Industrial Initiative (ESNII) will support three Generation IV reactor systems: a sodium-cooled fast reactor, or SFR, called Astrid that is proposed by France; a gas-cooled fast reactor, GFR, called Allegro supported by central and eastern Europe; and a lead-cooled fast reactor, LFR, technology pilot called Myrrha that is proposed by Belgium.". 
  10. ^ "The V4G4 Centre of Excellence for performing joint research, development and innovation in the field of Generation-4 (G4) nuclear reactors have been established. 20 July 2013 National Center for Nuclear Research (NCBJ]". 
  11. ^ "the European Gas cooled Fast Reactor.". 
  12. ^ "The GOFASTR research program". 
  13. ^ "Areva, CEA secure EUR650m funding to develop ASTRID sodium-cooled Generation IV reactor 11/11/2010". 
  14. ^ "UK and France Sign Landmark Civil Nuclear Cooperation Agreement 02/22/2012 . POWERnews". 
  15. ^ "Joint venture launched for Chinese fast reactor". 
  16. ^ Hellemans, Alexander (12 January 2012). "Reactor-Accelerator Hybrid Achieves Successful Test Run". Science Insider. Retrieved 29 December 2014. 
  17. ^ "Strategies to Address Global Warming". 
  18. ^ "4th Generation Nuclear Power". 
  19. ^ Warner, Ethan S.; Heath, Garvin A. Life Cycle Greenhouse Gas Emissions of Nuclear Electricity Generation: Systematic Review and Harmonization, Journal of Industrial Ecology, Yale University, published online April 17, 2012, doi: 10.1111/j.1530-9290.2012.00472.x.
  20. ^ Tabuchi, Hiroko (17 June 2011). "Japan Strains to Fix a Reactor Damaged Before Quake". The New York Times. 
  21. ^ a b "GIF Membership". Retrieved 28 October 2014. 
  22. ^ "generation IV international forum updates technology roadmap and builds future. DOE". 
  23. ^ "The Generation IV international forum holds their 36th meeting on Monday 18th Nov 2013 in Brussels.". 
  24. ^ Technology Roadmap Update for Generation IV Nuclear Energy Systems
  25. ^ Generation IV International Forum, overview, John E. Kelly, page 15

External links[edit]