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Type of nuclear reactor that operates at high temperatures as part of normal operation From Wikipedia, the free encyclopedia
A high-temperature gas-cooled reactor (HTGR) is a type of gas-cooled nuclear reactor which use uranium fuel and graphite moderation to produce very high reactor core output temperatures.[1] All existing HTGR reactors use helium coolant. The reactor core can be either a "prismatic block" (reminiscent of a conventional reactor core) or a "pebble-bed" core. China Huaneng Group currently operates HTR-PM, a 250 MW HTGR power plant in Shandong province, China.
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The high operating temperatures of HTGR reactors potentially enable applications such as process heat or hydrogen production via the thermochemical sulfur–iodine cycle. A proposed development of the HGTR is the Generation IV very-high-temperature reactor (VHTR) which would initially work with temperatures of 750 to 950 °C.
The use of a high-temperature, gas-cooled reactor for power production was proposed by in 1944 by Farrington Daniels, then associate director of the chemistry division at the University of Chicago's Metallurgical Laboratory. Initially, Daniels envisaged a reactor using beryllium moderator. Development of this high temperature design proposal continued at the Power Pile Division of the Clinton Laboratories (known now as Oak Ridge National Laboratory) until 1947.[2] Professor Rudolf Schulten in Germany also played a role in development during the 1950s. Peter Fortescue, whilst at General Atomics, was leader of the team responsible for the initial development of the High temperature gas-cooled reactor (HTGR), as well as the Gas-cooled fast reactor (GCFR) system.[3]
The Peach Bottom unit 1 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).[4][failed verification]
Experimental HTGRs have also existed in the United Kingdom (the Dragon reactor) and Germany (AVR reactor and THTR-300), and currently exist in Japan (the High-temperature engineering test reactor 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, the HTR-PM reactors, each with 100 MW of electrical production capacity, have gone operational in China as of 2021.[5]
The neutron moderator is graphite, although whether the reactor core is configured in graphite prismatic blocks or in graphite pebbles depends on the HTGR design.
The fuel used in HTGRs is coated fuel particles, such as TRISO[6][7][8][9] fuel particles. Coated fuel particles have fuel kernels, usually made of uranium dioxide, however, uranium carbide or uranium oxycarbide are also possibilities. Uranium oxycarbide combines uranium carbide with the uranium dioxide to reduce the oxygen stoichiometry. Less oxygen may lower the 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.[10] 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[11] concept conceived at Argonne National Laboratory has been used to better manage the excess of reactivity.
Helium has been the coolant used in all HTGRs to date. Helium is an inert gas, so it will generally not chemically react with any material.[12] Additionally, exposing helium to neutron radiation does not make it radioactive,[13] unlike most other possible coolants.
In the prismatic designs, control rods are 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.
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 high grade process heat. Reactors are designed for 60 years of service.[14]
As of 2011, a total of seven HTGR reactors had been constructed and operated.[15] A further two HTGR reactors were brought on-line at China's HTR-PM site, in 2021/22.
| Facility name |
Country | Commissioned | Shutdown | No. of reactors |
Fuel type | Outlet temperature (°C) |
Thermal power (MW) |
|---|---|---|---|---|---|---|---|
| Dragon reactor[15] | United Kingdom | 1965 | 1967 | 1 | Prismatic | 750 | 21.5 |
| Peach Bottom[15] | United States | 1967 | 1974 | 1 | Prismatic | 700–726 | 115 |
| AVR[15] | Germany | 1967 | 1988 | 1 | Pebble bed | 950 | 46 |
| Fort Saint Vrain[15] | United States | 1979 | 1989 | 1 | Prismatic | 777 | 842 |
| THTR-300[15] | Germany | 1985 | 1988 | 1 | Pebble bed | 750 | 750 |
| HTTR[15] | Japan | 1999 | Operational | 1 | Prismatic | 850–950 | 30 |
| HTR-10[15] | China | 2000 | Operational | 1 | Pebble bed | 700 | 10 |
| HTR-PM[16] | China | 2021 | Operational | 2 | Pebble bed | 750 | 250 |
Additionally, from 1969 to 1971, the 3 MW Ultra-High Temperature Reactor Experiment (UHTREX) was operated by Los Alamos National Laboratory to develop the technology of high-temperature gas-cooled reactors.[17] In UHTREX, unlike HTGR reactors, helium coolant contacted nuclear fuel directly, reaching temperatures in excess of 1300 °C.
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