Very high temperature reactor: Wikis


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Very high temperature reactor scheme.

The Very High Temperature Reactor is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. This reactor design envisions an outlet temperature of 1000°C. The reactor core can be either a “prismatic block” or a “pebble-bed” core. The high temperatures enable applications such as process heat or hydrogen production via the thermochemical sulfur-iodine cycle.

The reactor is intended for use in nuclear power plants to produce nuclear power from nuclear fuel.



AVR in Germany.

The earlier version of this design was known as a high temperature gas-cooled reactor or the HTGR, and to an extent, the general type of this reactor is still known by that name; the very high temperature reactor representing a modern and highly evolved version of the original HTGR design. The HTGR design was first proposed by the Staff of the Power Pile Division of the Clinton Laboratories (known now as Oak Ridge National Laboratory) in 1947.[1] Professor Dr. Rudolf Schulten in Germany also played a role in development during the 1950s. The Peach Bottom reactor in the United States was the first HTGR to produce electricity, and did so very successfully, with operation from 1966 through 1974 as a technology demonstrator. Fort St. Vrain Generating Station was one example of this design that operated as an HTGR from 1979 to 1989; though the reactor was beset by some problems which led to its decommissioning due to economic factors, it served as proof of the HTGR concept in the United States (though no new commercial HTGRs have been developed there since).[2] HTGRs have also existed in the United Kingdom (the Dragon reactor) and Germany (AVR and THTR-300), and currently exist in Japan (the HTTR using prismatic fuel with 30 MWth of capacity) and China (the HTR-10, a pebble-bed design with 10 MWe of generation). Two full-scale pebble-bed HTGRs, each with 100 - 195 MWe of electrical production capacity are under construction in China at the present as of November 2009[3], and are promoted in several countries by reactor designers.[4] More recently, this reactor design type has been substantially updated and is now proposed in a form known as the Very High Temperature Reactor in the United States.

Nuclear reactor design


Neutron moderator

Some United States and Russian designs refer to a prismatic block core configuration, where hexagonal graphite blocks are stacked to fit in a circular pressure vessel. Pebble bed designs are also being studied and have been used at lower temperatures than those envisioned for the VHTR. Pebble bed designs usually have a core where the pebbles are in an annulus, and there is a graphite center spire.

Nuclear fuel

The fuel is usually referenced to be uranium dioxide in a TRISO configuration, however, uranium carbide or uranium oxycarbide are also possibilities. A combination of uranium dioxide and uranium carbide may also be used to reduce internal pressure in the TRISO particles caused by the formation of carbon monoxide, due to the oxidization of the porous carbon layer in the particle.[5] The TRISO particles are either dispersed in a pebble for the pebble bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks. The QUADRISO fuel concept conceived at Argonne National Laboratory has been used to better manage the excess of reactivity.[6]



This helium cooled reactor type is the dominant one being studied; its primary design uses a 600-MW thermal core with a helium outlet temperature of 1,000°C. Helium has been used in most high temperature gas reactors (HTGR) to date. Helium is an inert gas, so it will not chemically react with any material.[7] Additionally, exposing helium to neutron radiation does not make it radioactive,[8] unlike most other possible coolants.

Molten salt

The molten salt cooled variant, the LS-VHTR, similar to the advanced high temperature reactor (AHTR) design, uses a liquid fluoride salt for cooling in a pebble core. It shares many features with a standard VHTR design, but uses molten salt as a coolant instead of helium. The pebble fuel floats in the salt, and thus pebbles are injected into the coolant flow to be carried to the bottom of the pebble bed, and are removed from the top of the bed for recirculation. The LS-VHTR has many attractive features, including: the ability to work at high temperatures (the boiling point of most molten salts being considered are >1,400°C), low pressure operation, high power density, better electric conversion efficiency than a helium-cooled VHTR operating at similar conditions, passive safety systems, and better retention of fission products in case an accident occurred.


In the prismatic designs, control rods would be inserted in holes cut in the graphite blocks that make up the core. The VHTR will be controlled like current PBMR designs if it utilizes a pebble bed core, the control rods will be inserted in the surrounding graphite reflector. Control can also be attained by adding pebbles containing neutron absorbers.

Materials Challenges

The high temperature, high neutron dose, and, if using a molten salt coolant, the corrosive environment[9], of the VHTR require materials that exceed the limitations of current nuclear reactors. In a study of Generation IV reactors in general (of which there are numerous designs, including the VHTR), Murty and Charit suggest that materials that have high dimensional stability, either with or without stress, maintain their tensile strength, ductility, creep resistance, etc after aging, and are corrosion resistant are primary candidates for use in VHTRs. Some materials suggested include nickel-base superalloys, silicon carbide, specific grades of graphite, high chromium steels, and refractory alloys.[10] Further research is being conducted at US national laboratories as to which specific issues must be addressed in the Generation IV VHTR prior to construction.

Safety features and other benefits

The design takes advantage of the inherent safety characteristics of a helium-cooled, graphite-moderated core with specific design optimizations. The graphite has large thermal inertia and the helium coolant is single phase, inert, and has no reactivity effects. The core is composed of graphite, has a high heat capacity and structural stability even at high temperatures. The fuel is coated uranium-oxycarbide which permits high burn-up (approaching 200 GWd/t) and retains fission products. The high average core-exit temperature of the VHTR (1,000°C) permits emissions-free production of process heat.

See also


  1. ^ McCullough, C. Rodgers; Staff, Power Pile Division (1947-09-15). "Summary Report on Design and Development of High Temperature Gas-Cooled Power Pile" (in English). Oak Ridge, TN, USA: Clinton Laboratories (now Oak Ridge National Laboratory). Retrieved 2009-11-23.  
  2. ^ IAEA HTGR Knowledge Base
  3. ^ Current status and technical description of Chinese 2 x 250 MWth HTR-PM demonstration plant
  4. ^ HTGR - High Temperature Gas-cooled Reactor _ Nuclear Pictures -
  5. ^ D. Olander J. Nucl. Mater. 389 (2009) 1-22.
  6. ^ Technical Articles on QUADRISO Fuel
  7. ^ "High temperature gas cool reactor technology development" (PDF). IAEA. 1996-11-15. pp. 61. Retrieved 2009-05-08.  
  8. ^ "Thermal performance and flow instabilities in a multi-channel, helium-cooled, porous metal divertor module". Inist. 2000. Retrieved 2009-05-08.  
  9. ^ D. T. Ingersoll, C. W. Forsberg, P. E. MacDonald, ORNL Technical Document, Oak Ridge Tennessee, ORNL/TM-2006/140, 2007, pp. 46
  10. ^ K.L. Murty, I. Charit, J. Nucl. Mater. 383 (2008) 189-195.

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