Folks,
This is a continuation in our series on nuclear reactors. We have already discussed
current nuclear reactor designs,
new nuclear reactor designs and
advanced nuclear reactor designs. In this post we will discuss Generation IV nuclear reactor designs. Material from the US DOE and Wikipedia has been freely used.
The reason for this focus on nuclear energy (aside from the fact that the reader's blogmeister works in the nuclear power industry) is to emphasize that we do not have an energy crisis nor a lack of obtaining fuel to provide electrical power and to generate hydrogen that can replace fossil fuel in motor vehicles. Rather, we have a crisis of human greed run amuk - fossil fuel companies in bed with government politicians in an orgy of profit seeking and power mongering via corporate socialism (also called croney capitalism). Corporate executive and government politician alike conspire in tightening the strangle-hold of addiction to coal, oil and "natural" gas while holding up the false promise of deliverance by shiney tin-foil solar cells or the twirling blade of wind mills, both of which by the nature of the inconstancy of weather necessitate spinning reserve from power plants that combust solid mineral rock (i.e., coal), liquid slime (i.e., oil) or the gaseous refuse of million-year old plant decay (i.e., "natural" gas). God has endowed enough thorium and uranium in Earth's crust to fuel all seven billion people on the planet at a level required to meet the consumption needs of the average American for tens of thousands of years. With access to low cost, non-polluting energy, no human being need ever starve, be homeless, or suffer from cold or excessive heat again. The reason why we do not do this is quite simply because of the love of money which St. Paul tells us in 1st Timothy 6:10 is the root of all evil. God has given us all the tools we need, and all we need is a sense of ethics and morality to use them. Nevertheless, before I am distracted onto another side-track, let us continue with the main thrust to today's entry on nuclear energy.
The evolution to Generation IV reactor designs is depicted in the diagram below.
Generation I in the 1950s and 60s in the US consisted of the Shippingport PWR and the Dresden BWR, and in Great Britain consisted of the Magnox gas cooled reactors.
Generation II in the 1970s to the present consisted of PWRs by Westinghouse, Combustion Engineering and Babcock and Wilcox, BWRs by GE and heavy water reactors in Canada called Candus.
Generation III reactor designs consisted of the Westinghouse AP600 PWR, the Combustion Engineering System 80+ PWR and advanced Candu heavy water reactors.
Generation III+ reactors designs include the GE-Hitachi ABWR and ESBWR, the Westinghouse AP1000, the Areva EPR and the Mitsubishi APWR, all of which are in various stages of design apporval before the US NRC.
The evolution to and a description of the designs for Generation IV reactors may be found in the slide show at the following hyperlinked title:
Introduction to Generation IV Nuclear Energy Systems and the International Forum
A brief description of each design follows:
Gas-cooled fast reactor features a fast-neutron-spectrum, helium-cooled reactor and closed fuel cycle
The high outlet temperature of the helium coolant used in the GFR system makes it possible to deliver electricity, hydrogen, or process heat with high efficiency. The reference reactor is a 1200-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.
It is anticipated that GFR systems will minimise the production of long-lived radioactive waste and make it possible to utilize fissile and fertile materials (including depleted uranium) two orders of magnitude more efficiently than thermal spectrum systems. The key challenges associated with this system concern the development of new fuels and materials capable of operating at temperatures of 850°C, the core design and helium turbine. The innovative GFR technologies and design features are intended to overcome the consequences of using a high-pressure gas with poor thermal characteristics to cool down a core with a low thermal inertia during depressurization events.
Very-high-temperature reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle.
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, Republic of 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.
The VHTR offers the potential for the cogeneration of electricity and hydrogen, alongside process heat applications. As the basic technology for VHTR systems has already been established in high temperature gas reactor plants, the design is an evolutionary development. However, the system’s aim of operating above 1000°C presents significant challenges in terms of fuel and materials development, as well as safety under transient conditions.
Supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water.
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.
As the system uses existing light water reactor technology, there is already extensive worldwide experience in constructing and operating this sort of reactor. Proposed designs are likely to enjoy high thermal efficiency and a simplified system configuration. A SCWR design could be developed with a fast neutron spectrum. Using fast neutrons with higher kinetic energies would enable the system to produce at least as much fissile material as it consumes (thereby fulfilling the sustainability goal as set out in the Generation IV roadmap). This concept’s tendency to have a positive void reactivity coefficient together with the potential for design basis loss-of-coolant accidents are likely to make this difficult to develop. The other major challenges for the SCWR are to develop a viable core design, accurately estimate the heat transfer coefficient and develop materials for the fuel and core structure that will be sufficiently corrosion-resistant to withstand SCWR conditions.
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 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 power conversion system. Water/steam and carbon-dioxide are being considered as the working fluids for the power conversion system in order to achieve high-level performances in thermal efficiency, safety and reliability. With innovations to reduce capital cost, the SFR can serve markets for electricity.
The fuel cycle employs a full actinide recycle with three major options. The first option is a large size (600 to 1,500 MWe) loop-type sodium-cooled reactor using 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 second option is an intermediate size (300 to 600 MWe) pool-type reactor and the third a small size (50 to 150MWe) modular-type sodium-cooled reactor employing uranium-plutonium-minor-actinide-zirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in facilities integrated with the reactor. The outlet temperature is approximately 550 degrees celsius for all the three concepts.
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.
The SFR system already benefits from considerable technological experience, but also offers the potential to operate with a high conversion fast spectrum core, with the resulting benefit of increasing the utilization of fuel resources. The envisaged SFR capability to efficiently and nearly completely consume trans-uranium as fuel would reduce the actinide loadings in the high-level radioactive waste it produces. Such reductions would bring benefits in the radioactive waste disposal requirements associated with the system and enhance its non-proliferation attributes. Reducing the capital cost and improving passive safety, especially under transient conditions, are the major challenges for the SFR system.
Lead-cooled fast reactor (LFR) features a fast-spectrum, lead/bismuth eutectic liquid-metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.
The LFR system has excellent materials management capabilities since it operates in the fast-neutron spectrum and uses a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner to consume actinides from spent LWR fuel and as a burner/breeder with thorium matrices. An important feature of the LFR is the enhanced safety that results from the choice of molten lead as a relatively inert coolant. In terms of sustainability, lead is abundant and hence available, even in case of deployment of a large number of reactors. More importantly, as with other fast systems, fuel sustainability is greatly enhanced by the conversion capabilities of the LFR fuel cycle.
The LFR was primarily envisioned for missions in electricity and hydrogen production, and actinide management. Given its R and D needs in the areas of fuels, materials, and corrosion control, a two step process leading to industrial deployment of the LFR system has been envisioned: by 2025 for reactors operating with relatively low primary coolant temperature and low power density; and by 2035 for more advanced designs. The preliminary evaluation of the LFR concepts considered by the LFR Provisional System Steering Committee (PSSC) covers their performance in the areas of sustainability, economics, safety and reliability and proliferation resistance and physical protection.
The LFR concepts that are currently being designed are two pool-type reactors:
(1) the Small Secure Transportable Autonomous Reactor (SSTAR), developed in the USA and
(2) the European Lead-cooled System (ELSY), developed by the EC.
The main advantages of the LFR system are its expected fuel efficiency, its capabilities in terms of nuclear materials management (thereby mitigating proliferation risks) and the reduced production of high-level radioactive waste and actinides.
Molten salt reactor (MSR) produces fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full actinide recycling fuel cycle.

In a Molten Salt Reactor (MSR), the fuel is dissolved in a fluoride salt coolant. Prior MSRs were mainly considered as thermal-neutron-spectrum graphite-moderated concepts. Since 2005 R&D has focused on the development of fast-spectrum MSR concepts (MSFR) combining the generic assets of fast neutron reactors (extended resource utilization, waste minimization) to those relating to molten salt fluorides as fluid fuel and coolant (favourable thermal-hydraulic properties, high boiling temperature, optical transparency). In addition, MSFRs exhibit large negative temperature and void reactivity coefficients, a unique safety characteristic not found in solid-fuel fast reactors. MSFR systems have been recognized as a long term alternative to solid-fuelled fast neutron systems with unique potential (negative feedback coefficients, smaller fissile inventory, easy in-service inspection, simplified fuel cycle, etc.).
Apart from MSR systems, other advanced reactor concepts are being studied employing liquid salt technology as primary coolant in Fluoride-cooled High-temperature Reactor (FHR), or intermediate coolant as an alternative to secondary sodium in Sodium Fast Reactors (SFR) and to intermediate helium in Very High Temperature Reactors (VHTR).
More generally speaking, the development of higher temperature salts as coolants could bring new nuclear and non-nuclear applications. These salts could facilitate heat transfer for nuclear hydrogen production concepts, concentrated solar electricity generation, oil refineries and shale oil processing facilities amongst other applications.
Fluoride-cooled High-temperature Reactors (FHRs) combine the use of liquid fluoride salt coolants (like MSRs), pool type cores and vessel configurations in common with many sodium reactor designs, and coated particle fuels similar to high temperature gas-cooled reactors (Forsberg et al., 2008). The two most developed FHR designs are the 1200 MWe Advanced High Temperature Reactor (AHTR) that employs prismatic fuel elements and the 410MWe Pebble Bed Advanced High Temperature Reactor (PB-AHTR). The better fluoride salt heat transport characteristics, as compared to helium, enable power densities 4 to 8 times greater as well as power levels over 4000 MWt with passive safety systems. Fuel cycle characteristics are essentially identical to those of the VHTR, while intermediate heat transport, power conversion and balance of plant are essentially identical to those of the “reference” MSR.