Ahigh-temperature gas-cooled reactor (HTGR) is a type ofgas-cooled nuclear reactor which uses uranium fuel andgraphite moderation to produce very highreactor core output temperatures.[1] All existing HTGR reactors usehelium 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 operatesHTR-PM, a 250 MW HTGR power plant with two pebble-bed HTGRs, inShandong province,China.
The high operating temperatures of HTGR reactors potentially enable applications such as process heat orhydrogen production via the thermochemicalsulfur–iodine cycle. A proposed development of the HTGR is theGeneration IVvery-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 byFarrington Daniels, then associate director of the chemistry division at the University of Chicago'sMetallurgical Laboratory. Initially, Daniels envisaged a reactor usingberyllium moderator. Development of this high temperature design proposal continued at the Power Pile Division of the Clinton Laboratories (known now asOak Ridge National Laboratory) until 1947.[2] ProfessorRudolf Schulten inGermany also played a role in development during the 1950s.Peter Fortescue, whilst atGeneral Atomics, was leader of the team responsible for the initial development of the High temperature gas-cooled reactor (HTGR), as well as theGas-cooled fast reactor (GCFR) system.[3]
The United States'Peach Bottom Unit 1 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 (theDragon reactor) and Germany (AVR reactor andTHTR-300), and currently exist in Japan (theHigh-temperature engineering test reactor using prismatic fuel with 30MWth of capacity) and China (theHTR-10, a pebble-bed design with 10 MWe of generation). Two full-scale pebble-bed HTGRs, theHTR-PM reactors, each with 100 MW of electrical production capacity, have gone operational in China as of 2021.[5]
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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 asTRISO[6] fuel particles. Coated fuel particles have fuel kernels, usually made ofuranium 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.[7] 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[8] concept conceived atArgonne 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 aninert gas, so it will generally not chemically react with any material.[9] Additionally, exposing helium to neutron radiation does not make it radioactive,[10] 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 currentPBMR designs if it utilizes a pebble bed core, the control rods will be inserted in the surrounding graphitereflector. Control can also be attained by adding pebbles containingneutron absorbers.
The design takes advantage of the inherent safety characteristics of a helium-cooled, graphite-moderated core with specific design optimizations. The graphite has largethermal 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 gradeprocess heat. Reactors are designed for 60 years of service.[11]
As of 2011, a total of seven HTGR reactors have been constructed and operated.[12] A further two HTGR reactors were brought on-line at China'sHTR-PM site, in 2021/22.
| Facility name | Country | Commissioned | Shutdown | No. of reactors | Fuel type | Outlet temperature (°C) | Thermal power (MW) |
|---|---|---|---|---|---|---|---|
| Dragon reactor[12] | United Kingdom | 1965 | 1976 | 1 | Prismatic | 750 | 21.5 |
| Peach Bottom[12] | United States | 1967 | 1974 | 1 | Prismatic | 700–726 | 115 |
| AVR[12] | Germany | 1967 | 1988 | 1 | Pebble bed | 950 | 46 |
| Fort Saint Vrain[12] | United States | 1979 | 1989 | 1 | Prismatic | 777 | 842 |
| THTR-300[12] | Germany | 1985 | 1988 | 1 | Pebble bed | 750 | 750 |
| HTTR[12] | Japan | 1999 | Operational | 1 | Prismatic | 850–950 | 30 |
| HTR-10[12] | China | 2000 | Operational | 1 | Pebble bed | 700 | 10 |
| HTR-PM[13] | China | 2021 | Operational | 2 | Pebble bed | 750 | 250 |
| HTGR (Lianyungang)[14] | China | Approved for construction | 1 | ? | 660 |
Additionally, from 1969 to 1971, the 3 MWUltra-High Temperature Reactor Experiment (UHTREX) was operated byLos Alamos National Laboratory to develop the technology of high-temperature gas-cooled reactors.[15] In UHTREX, unlike HTGR reactors, helium coolant contacted nuclear fuel directly, reaching temperatures in excess of 1300 °C.
The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures
To date, seven HTGR plants have been built and operated
Three of these plants, AVR, Peach Bottom, and Fort St. Vrain, are actual electrical generating plants, and two, Dragon and UHTREX, are experimental plants being used primarily to develop the technology of high – temperature, gas-cooled reactors.
