Title: The High Temperature Gas-cooled Reactor: Safety considerations of the (V)HTR-Modul
Authors: KUGELER KURTNABIELEK HEINZBUCKTHORPE DEREK
Editors: SCHEUERMANN WALTER
HANEKLAUS NILS
FUETTERER MICHAEL
Publisher: Publications Office of the European Union
Publication Year: 2017
JRC N°: JRC107642
ISBN: 978-92-79-71312-5 (print)
978-92-79-71311-8 (online)
ISSN: 1018-5593 (print)
1831-9424 (online)
Other Identifiers: EUR 28712 EN
OP KJ-NA-28712-EN-C (print)
OP KJ-NA-28712-EN-N (online)
URI: http://publications.jrc.ec.europa.eu/repository/handle/JRC107642
DOI: 10.2760/270321
10.2760/970340
Type: Books
Abstract: Nuclear energy production is a recent technology. The first nuclear reactor demonstrating the feasibility of a sustained and controlled chain reaction was built in Chicago in 1942. In 1957, a nuclear reactor produced electricity for the first time. For the past 60 years, since then a significant technological progress has been achieved. Three generations of nuclear reactors have been successively developed and a fourth is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. The technology is mature, with approximately 450 nuclear reactors currently providing 17% of the world’s electricity, without greenhouse gas emissions. It must also be stressed that the technological progress and innovations achieved or currently being developed are all based on the same fundamental physical principles of nuclear fission, whose feasibility was demonstrated some 70 years ago. Heavy nuclides, like uranium, plutonium or thorium are fissioned by neutrons and release heat within the fuel material confining the radioactivity. This heat is extracted while simultaneously cooling the fuel by circulating a coolant (water in current European light water reactors). The heat recovered is used to run a turbine and a generator, which produces electricity. A lot of effort has been invested in international cooperation to define goals for the nuclear energy systems of the future, and also select the key technologies for achieving them. The effort has been made primarily through the Generation IV International Forum (GIF) that the American Department of Energy has initiated in 2000. The technology roadmap, developed by GIF is the starting point to identify and organize research and development of the new generation of reactors to be built around 2030 to 2040. To do so, 4 goals have been defined. Each of the systems comprises a nuclear reactor, an energy conversion system and the necessary fuel cycle, fuel manufacturing, spent fuel and final waste management equipment. The 4 major goals: • Sustainability • Economics • Safety and reliability • Proliferation Resistance and physical protection have been separated out into fifteen criteria and twenty-four performance indicators or metrics. A final selection criterion was the degree of technological innovation in the candidate systems – which provides an excellent grounds for wide-reaching international cooperation – and the possible spin-offs for the other nuclear systems, or for the current or next generation of nuclear reactors. The following six systems deemed the most promising at the end of this evaluation exercise were called on to rally Forum cooperation on development work starting from 2004: • VHTR - Very High-Temperature Reactor system, over 1,000 °C, helium-cooled, dedicated to hydrogen production or hydrogen/electricity cogeneration; • GFR - Gas-cooled Fast Reactor system – Helium-cooled fast reactor; • SFR - Sodium-cooled Fast Reactor system; • SCWR - SuperCritical Water-cooled Reactor system; • LFR - Lead-cooled Fast Reactor system – Lead or Pb-Bi alloy-cooled Fast reactor; • MSR - Molten Salt Reactor. VHTR scored very well with the exception concerning sustainability because of its open fuel cycle, which requires reprocessing to increase sustainability. From a technological point of view the VHTR is a further development of the High Temperature Reactor (HTR) which has been developed in the years from 1960 to 2001. Two basic fuel element designs of the HTR have been developed and implemented, one in Germany, the other in USA. The main characteristic of the German design is the conditioning of compacted microparticles in a graphite matrix in the form of 60-mm diameter spheres. They are continuously inserted in and extracted from the reactor at a rate of one pebble approximately every 20 seconds. When a pebble reaches its maximum depletion rate, it is replaced with a new one. The first German HTR was the AVR built at the Jülich research centre. This centre has maintained very high competence in this technology. The AVR set new records (for HTRs) in terms of performance and operating duration. Its construction began in 1961. It was connected to the electric power network in 1966 (15 MWel) and shut down in 1988. It served as an experimental platform for fuel technology development within the scope of cooperation between the Jülich research centre and NUKEM, the fuel industrial manufacturer still considered as a reference today. The core temperature of 850 °C at the start of operation was increased to 950 °C. The steel pressure vessel design served to achieve design transients (e.g., core cooling loss) that contributed to validating the safety concepts applied in this type of reactor. The AVR showed the viability of the pebble bed concept and demonstrated its reliability through physical tests for which the plant was not initially designed. A loss of coolant flow without scram was simulated in 1970, and a loss of coolant transient was also achieved prior to final shutdown. The fuel has undergone significant developments and improvements at the Jülich research centre, in partnership with NUKEM (manufacturer). The second HTR built in Germany was the THTR-300 (Thorium High-Temperature Reactor), which went critical in 1983. This was a 300 MWel commercial reactor with a concrete vessel, built by Brown Boveri. The operation of the THTR was marked by a number of technical problems that did not seem impossible to overcome. In particular, a planned inspection in 1988 revealed the rupture of a number of bolts securing hot duct insulation plates which, combined with an unfavourable political context, led to the decision to permanently decommission the facility in 1989 after only 423 equivalent full power days The American family differs from the German family mainly in terms of core and fuel organization. The core is composed of prismatic graphite blocks containing the fuel compacts. The first commercial implementation was Peach Bottom (40 MWel), which went critical in early 1966 and was shut down in 1974. After the discovery of an increasing number of fuel cladding ruptures, a second core was fabricated using a more advanced technology improving the quality of the first porous graphite layer and the characteristics of successive layers. 93% availability was achieved during irradiation of this second core, and reactor coolant system activity remained extremely low, indicating the excellent quality of the new fuel. The reactor subsequently operated without major problems and was shut down for economic reasons. The second implementation was Fort Saint-Vrain (330 MWel), whose construction began in 1968 and which went critical in 1974. Its operation was marked by technical problems (namely accidental water ingress in the reactor coolant system causing accelerated corrosion of steel components and poor availability) and it was finally decommissioned in 1989. Despite the negative functional aspects of its operation, the excellent leak-tightness of the fuel elements led to very positive radiological results for operation and maintenance activities, with the exception of tritium releases due to water leaks. On the whole, the operating experience from German and American prototypes has largely confirmed the technical expectations regarding HTRs, i.e.: • Very good behaviour of the particle fuel under irradiation, even at high temperatures, and low release of fission products in the coolant gas providing very clean reactors; • Possibility of using high-temperature helium as coolant gas; • Easy control, high thermal inertia and significant operating safety margins (demonstrated at real scale with the AVR). The impact of the Three Mile Island accident and the excellent safety characteristics of HTRs (thermal inertia, good apparent core conductivity, low power density) have led to research on configurations allowing completely passive residual power evacuation. Low-power HTRs are particularly well suited to satisfy this new passive safety requirement, like the concept of the German HTR Modular reactor which has been developed by Siemens-Interatom. This 80 MWel reactor uses the radiating capacity of a metal vessel to ensure passive cooling of the fuel, whose temperature remains below 1600 °C regardless of accident conditions. This reactor was never build, but a detailed design has been completed and some licensing issues have been solved. HTR reactors like the HTR Modul with an electric power of 100 to 300 MWel set the trend of current design, whilst in current European research projects cogeneration is in favour of electricity generation.
JRC Directorate:Nuclear Safety and Security

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