High-Temperature Gas-Cooled Reactors

High-temperature gas-cooled reactors (HTGRs), also known as very-high-temperature reactors (VHTR) are Generation IV reactors that can operate at very high temperatures and use a graphite-moderated gas-cooled nuclear reactor with a once-through uranium fuel cycle.

This design permits a very high outlet temperature in the order of 1 000°C. The first HTGR design was proposed at the Clinton Laboratories (now Oak Ridge National Laboratory) in 1947. Germany also played a significant role in HTGR development over the next decade. The Peach Bottom reactor in the United States (US) was the first HTGR to produce electricity, with operation from 1966 through 1974 as a 150 MW(th) demonstration plant. The Fort St. Vrain plant was the first commercial power design, operating from 1979 to 1989 with a power rating of 842 MW(th). Even though the reactor was beset by operational issues that led to its decommissioning due to economic factors, it served as proof of the HTGR concept in the United States. 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 the People’s Republic of 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, and are promoted in several countries by reactor designers. The US Department of Energy (DOE) Next Generation Nuclear Plant (NGNP) represents a significant and growing activity in the United States.

Publications and reports
NEA work on high-temperature gas-cooled reactors

In response to increasing interest in HTGRs and the need for improved knowledge of materials for nuclear applications that resist high temperatures, the NEA organised a series of three information exchange meetings on basic studies in the field of high-temperature engineering. 

  1. First meeting proceedings (1999) cover studies on irradiation effects on advanced materials, safety-related behaviour of HTGRs and in-pile reactor instrumentation development. They also include recommendations for further promotion of international collaboration.
  2. Second meeting proceedings (2001) provide an overview of the activities being carried out in eight countries, the improvement of material properties for HTGR application, in-core monitoring methods and properties of irradiated graphite, and HTGR fuel fabrication and performance.
  3. Third meeting proceedings (2003) provide a summary of high-temperature research currently under way, including studies on the behaviour of irradiated graphite and improvements in material properties under high-temperature irradiation. They also contain recommendations for further international work in the areas of high-temperature engineering.


Ian Hill