Generation IV Nuclear Reactor Comparisons

Common energy plants (coal, natural gas, nuclear) employ a Rankine steam cycle for electric power production. Specifically, nuclear plants using the Rankine cycle are limited by the amount of heat they can produce, radioactive embrittlement, and radiated coolant. Mentioned below are a few considerations for a new generation of nuclear power. Designs intend to increase efficiency, increase safety and reliability, and allow the opportunity for other commercial benefits, such as hydrogen production.

Very-High-Temperature Reactor (VHTR)



The typical VHTR design employs a helium coolant operated under the Brayton power cycle with the intension of achieving working temperatures in excess of 1000°C. The initial design was developed in the 1980's and has proven its capabilities in several countries. South Africa's pebble bed modular reactor (PBMR) is one such design that offers several advantageous characteristics. Thorium-232, in conjuction with Uranium-235, is encased in a TRISO coating and modulated by a graphite shell. The fuel design allows constant recycling/reprocessing/refueling of the reactor core; eliminating a significant percentage of shut-down losses associated with rod lattice designs.

The helium gas provides a nonreactive cooling medium with a low nuclear cross section that can either directly drive a turbine for power production, or can be directed to process heat to increase the efficiency of hydrogen gas production to an economical level. The gas coolant adds another level of redundancy to the safety of the system. In the event of fuel cladding failure, radioactivity will be contained within the core without the possibility of transport via coolant.

Other coolants have been considered, but have not been tested. Molten salt (mentioned in detail below) offers the advantage of very high boiling temperatures with high radioactivity retention potential, but restricts visual inspection of the core and is chemically highly reactive.


• Helium Coolant
• Thorium and Uranium Fuel
• Higher electric power efficiencies (>50%)1
• Hydrogen production


• Fuel may not be rotated out when it needs to be
• Low thermal conductivity of Helium
• Limited research

Super Critical Water Cooled Reactor (SWCR)



This reactor uses water as the primary coolant. The cycle is a once through design, and is basically a LWR operating at higher pressures and temperatures. Because the water is at such a high pressure, it will never experience a phase change when heating up or cooling off. Its main purpose is to provide low cost electricity generation. There is a great interest in this design because it builds on the existing technologies driving current PWR and BWR designs.


As the water leaves this reactor core, it will be 500° C. As one can see in the illustration above, because the water is incredibly hot, and at such high pressures, it will never actually boil. This means the water will not change phase. The fuel that will be responsible for reacting and heating the water is a low-enriched uranium.

The cycle that this design uses really is very simplistic. It has no need for steam generators or steam evaporators like its PWR, BWR cousins. It has no recirculation pump, and has half the number of steam lines. One of the other greatest aspects about this plant is the incredible small size of the reactor itself, smaller than both the PWR and BWR.


• 45% thermal efficiency
• Able to use majority of PWR and BWR reactor vessel components
• No need for jet pumps, pressurizers, or dryers
• Next logical improvements to the PWR and BWR configurations


• Great deal of research needs to be done on super critical water
• High pressure of water creates greater chance of loss of coolant accident (LOCA)
• High pressure and temperatures require new analysis of capable pipes and components

Molten Salt Reactor (MSR)



Taking advantage of technology available since the 1960's, Molten Salt Reactors have been designed for a plethora of uses. From commercial plants, to nuclear powered bomber aircraft, the MSR has the advantage of low pressure operation with higher core heat transfer. This allows for a reduced reactor size with fewer pumps and pipes, operating at higher efficiencies. There are two proposals for the Generation IV MSR designs, molten salt fueled reactors and molten salt cooled reactors.

Molten salt fueled reactors are not as suceptible to corrosion concerns (uranium and thorium reduce chemical reactivity when bonded to flourine). Low operating pressures allow the possibility for multiple small cores rather than one large reactor. Because the fuel is dissolved within the salt, there is no costs for fuel fabrication. Also, the potential for breeding is available. Molten salt cooled reactors offer the same advantages of low operating pressures and the inherit safety of containing fission products, while reducing the concern of a steam explosion; but the corrosion resistance is all but eliminated by removing the fuel from the coolant. Materials suitable for containing the corrosive environment at typical reactor temperatures have not been developed.

The chemical characteristics of molten salts demand constant reprocessing and purification. Flouride salts react with water, creating hyrdoflouric acid, which is incredibly corrosive. The reprocessing is advantageous in that it removes fission products, increasing the neutron economy of the reactor. The safety advantages (retention of fission products, lower risk of explosion, less risk of departure from nucleate boiling), combined with the higher efficiencies associated with higher operating temperatures, encourages the new design proposals.


• Allows for small reactor size
• Technology is researched and proven
• Higher operating temperatures
• Chemical retention of fission products


• High corrosion potential
• Unknown material required for corrosion resistance

Gas-Cooled Fast Reactor (GFR)



Similar to the VHTR, the GRF uses helium coolant operating a brayton power cycle to generate electricity. The helium offers the same advantages and reliability mentioned above. The advantaged of the GFR is its breeding capabilities. Fertile uranium, as well as several other fissile fuels, can be used without the need for neutron moderation. Due to the absorption properties of the fuels responsive to the fast neutron spectrum, fuel can be produced in the reactor over time, ultimately creating more fuel than what was originally installed in the core.

Core redesign, to accommodate a high temperature accident and fast neutron damage, has advantages and disadvantages. Unconventional cladding material would need to be tested and proven to withstand the failure criteria and environment, which would undoubtedly lengthen construction plans. Different cladding, such as Zr3Si2 has the benefit of increased heat transfer with fast neutron reflection characteristics not found in current clad design.


• Breeding capabilities
• Higher power density
• New fuel design


• Low thermal conductivity of Helium
• Fast neutron damage
• Limited research

Sodium Cooled Fast Reactor (SFR)



This generation 4 reactor usually comes in 2 different sizes according to the energy output desired. The smaller size will accommodate 150 to 600 MWe while the larger size handles 500 to 1,500 MWe. Both of these setups use sodium as the moderator. Sodium is a heavier material, when neutrons collide with sodium atoms, they do not lose as much energy as with water. This is a main advantage to using this kind of reactor. Both of these reactors use uranium-plutonium mixed fuels. The small unit uses the fuel mixed with metal alloys while the larger uses the uranium-plutonium oxide.

The unique aspects of using sodium within a fast reactor are the energies able to be removed from the reactions from the radioactive materials as well as the safety features that come from using a metal coolant rather than water (Those will be detailed later). The sodium typically leaves the reactor unpressurized at 510° to 550° C. Usually, the designs for the SFR favors a coolant pool as illustrated in the picture above. After the coolant leaves the reactor, it moves along to a heat exchanger where the heat energy is transferred to another loop in the power plant. This loop travels to the turbines to provide the mechanical energy for the generator.

This reactor’s development is mainly determined by its fuel material developments. It is said to be the most realizable Generation 4 reactor and could be used for electricity production in the USA in about 12 years time. These reactors are the most developed of all the Generation 4 reactors. Several units have been built and are in operation around Europe.


• Advantages of every other fast reactor
• 2 primary fuel cycle technologies
• Recovers and recycles 99.9% of the actinides
• Inherently low decontamination factor of the product
• Never separates plutonium at any stage
• Achieves thermal efficiency of 40%


• Expensive
• Fuel system R&D still needed
• Overall system research is still needed to verify passive safety systems and component design
• Sodium catches fire and explodes when in contact with water or air.

Lead Cooled Fast Reactor (LRF)



This fourth generation nuclear reactor uses lead or lead/bismuth as the primary coolant for within the reactor core. It makes use of the fast neutron spectrum and a closed fuel cycle. The unit also is capable of the Multi-TRU recycle process. Small plants can be designed to handle 50 to 150 MWe, medium sized covering 300 to 400 MWe and even a unit that will generate 1,200 MWe. This plant has been designed as a modular configuration, where the components making up the plants can me manufactured off site and brought in and pieced together. Each design is rated for anywhere from 15 through 30 years of operation before any kind of reevaluation or modification upgrade is done.

The fuel that is used is uranium with either metal or nitride. The lead leaves the reactor core at about 550° to 800° C. The lead alloy has a low neutron absorption and slow down power which facilitates natural circulation. The primary coolant loop operates unpressurized which allows future designs employing passive safety.

This unit has been designed specifically for the electricity utilities of the world. It can provide cheap and reliable energy because it will either be self sufficient or can employ the TRU process for refueling. The system also employs a range of new technologies within the plant, they are: natural circulation, lift pumps, direct contact heat exchangers and direct contact steam generators.


• Operates unpressurized removing potential LOCA's
• Design can be manufactured off site and assembled where needed
• Design can be easily modified to operate with H2
• Long refueling interval, 10 to 20 years


• Requires a great deal of research and development to become mainstream
• New fuel needs to be analyzed for performance specification
• Component design places new risks on outside manufacturers
• Will probably require 22 more years of research before commercially available

: The Sodium Cooled Fast Reactor : M.J. Lineberry and T.R. Allen, Argonne National Laboratory : full source reference : Supercritical Water Cooled Reactor System as one of the most promising type of Generation 4 Nuclear Reactor Systems : David Danielyan, November 24, 2004 : Lead Cooled Fast Reactor : T.R. Allen and D.C. Wade, American Nuclear Society Winter Meeting, November 18 2002 : The Gas-Cooled Fast Reactor System, Hussein Khalil
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