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1194 the Evolution and Future Development of the High GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1194 The Evolution and Future Development of the High Temperature Gas Cooled Reactor H. L. Brey* Consultant to the Japan Atomic Energy Research Institute This report chronologically traces the development of the high temperature gas cooled reactor (HTGR) and includes an examination of current international research activities and plant designs. The evolution of the HTGR as a safe, environmentally agreeable and efficient energy source for the generation of electricity and for a broad range of high temperature industrial heat applications has been on-going throughout the past four decades. This report depicts the development of the HTGR through international plant designs and selected facilities from initial research to the present focus on the modular gas turbine and high temperature heat applications such as the production of hydrogen 1,2). This advanced nuclear heat process, with an average core outlet temperature approaching 950°C and using ceramic coated fuel in a helium cooled, graphite matrix, provides a broad and unique set of features and physical properties for scientific investigation as a major energy source for supporting the needs of society well into the 21st century. This paper provides a review of current HTGR related R & D activities including an overview of the HTTR and HTR-10 test reactors, status of GT-MHR and PBMR plant development, the U.S. HTGR programme and a synopsis of the multi-national efforts of the EC and Generation IV nuclear power technology program. KEYWORDS: advanced reactor designs and development; gas cooled reactors I. Introduction Interest in the HTGR has resulted in a wide array of international research and development activities including the multi-national efforts of the EC, Generation The historical development of the HTGR is traced IV nuclear power technology and on-going R&D within through individual plant designs and selected facilities the Member States associated with the IAEA. from initial research and continuing through the design of the modular steam cycle plant. This historical path, with A tabulation of the design characteristics of many some discretionary latitude, can be divided into the of the HTGR plants, both past and future, is provided as following four general areas: Appendix A at the end of this report. • Early Gas Cooled Reactor (GCR) Development II. Historical Review of the HTGR • HTGR Prototype Plants • Demonstration Plants and Large Plant Designs 1. Early GCR Development • Modular HTR Steam Cycle Plant Development Gas-cooled reactor history effectively began with Current and projected HTGR development the startup in November 1943 of the graphite-moderated, activities that are anticipated to extend into the early air-cooled, 3.5-MW, X-10 reactor in Oak Ridge, st decades of the 21 century are divided into the following Tennessee. Commercial gas-cooled nuclear power began three general areas: in 1953 when the United Kingdom decided to combine plutonium production with electric power generation, and • HTGR Powered Process Heat Applications work was started on the four-unit power station at Calder • HTTR and HTR-10 Test Reactors Hall. • Modular HTGR Gas Turbine Plant Development 1,2) _______________________________________________ These first power reactors were graphite * Corresponding author, Tel. +1-970-240-2791, E-mail: moderated with natural-uranium metal rods and cooled by forced circulation of carbon dioxide at a pressure of 0.8 [email protected] MPa and at an outlet temperature of 335°C. The U.K.'s extensive commitment to GCR technology has included construction of 26 Magnox reactors and 14 advanced gas- The Dragon reactor in the U.K. was the first cooled reactors (AGRs), which deliver steam at the HTGR prototype. It began in 1959 as an international modern conditions of 538°C and 16.5MPa. project of the European Organization for Economic Cooperation and Development 5). The initial objective of France's early interest in the GCR aided this project was to demonstrate the feasibility of the HTGR development in the United Kingdom. In 1951, the 2-MW and to launch the development of a technology which had research reactor at Saclay, which began operating with already begun at a low level in various national nitrogen coolant and later switched to carbon dioxide, was laboratories. The Dragon Reactor Experiment first the first gas-cooled reactor to use closed circuit, operated at power in July 1965 and reached its full-power pressurized cooling. These experiments, coupled with the operation of 20MW(t) by April 1966. The reactor had a experience of the air-cooled, open-circuit, G1 plutonium steel pressure vessel, graphite fuel elements with high- production reactor, formed the basis for France's GCR enriched uranium-thorium carbide coated fuel particles, program. Although similar to the U.K's program in coolant, and a helium coolant. Exit helium at 750°C was circulated moderator, and fuel, the French program introduced the through primary heat exchangers and returned at 350°C use of the prestressed concrete reactor vessel (PCRV), (electric power generation was not a feature of this which the U. K. adopted for its later Magnox reactors and prototype). all of its AGRS. The reactor operated for long periods at full The first nuclear power reactor in Japan, which power and demonstrated the successful operation of many started commercial operation in July 1966, was the carbon components, e.g., the helium circuit purification system, dioxide-cooled, 166MW(e) Tokai station located 80 miles the control rod drives, the fuel handling equipment, the northeast of Tokyo. The plant design generally followed gas-bearing circulators, and the reactor fuel. As a test bed, the design of the U.K's Magnox reactors; however, it provided much information on fuel and material because of population concerns, its containment design irradiation tests and component tests under high-purity provided a partial third barrier for a postulated release of helium conditions. The reactor operated until March 1976, coolant by sealing after pressure decay. In 1969, the Japan at which time the project was terminated 6). Atomic Energy Research Institute (JAERI) initiated studies on the very high temperature gas-cooled reactor in Simultaneously with the initiation of the Dragon recognition that a nuclear process heat source of 900°C or project, interest in Germany and the U.S. led to the 15 higher would find use in coal gasification and hydrogen MW(e) AVR at Julich, Germany, and the 40MW(e) HTGR and methanol production 3). at Peach Bottom in the USA. Development of the HTGR began in the 1950s to (2) The AVR improve upon GCR performance. HTGRs utilize ceramic fuel particles surrounded by coatings and dispersed in a In August 1959, the order was placed with the graphite matrix, along with a graphite moderator. Either German partnership of BBC/Krupp for the construction of prismatic graphite moderator blocks or spherical fuel an experimental nuclear power plant with a high elements (pebbles) are employed. Helium is used as the temperature reactor of 46MW(t)/15 MW(e). Construction coolant to permit an increase in the operating temperature, work on the plant at the KFA Nuclear Research Center in and flows through coolant holes in the block type elements, Juelich began in 1961. Initial reactor criticality occurred in or through the interstices present in the pebble bed core. August 1966, and on 17 December 1967 electricity was HTGRs can operate at very high core outlet coolant supplied to the public supply grid for the first time. In temperatures because of the use of an all ceramic core 4). February 1974, the mean gas outlet temperature, which had initially been set at 850°C, was increased to 950°C. 2. HTGR Prototype Plants Operation of the AVR continued until December 1988. In completing 21 years of service, the plant had accumulated The initial HTGRs included the Dragon reactor more than 122,000 hours of operation with a 66.4% overall availability and had generated 1.67 billion kwhr of experiment, the Arbeitsgemeinshaft Versuchsreaktor 4) (AVR) and Peach Bottom (No. 1). Common among these electricity . plants were the utilization of steel vessels to contain the primary system and the attainment of high core outlet The design of the AVR included a steel temperatures culminating in the AVR achieving extended containment vessel and used particle fueled, graphite operation at 950°C. spheres 6 cm. in diameter that traveled downward through the core. Although the AVR initially included a core outlet temperature of 850°C, this was subsequently raised to (1) Dragon Reactor Experiment 950°C without decrease in plant performance. level. These two features were to be dominant throughout The AVR was the main fuel development tool for HTGR development of the 1970s and early 1980s. the pebble bed concept, and it and supplementary Although the general focus of the HTGR plant designers laboratory fuel testing became the major support of was on the development of large steam cycle units, Germany’s position that an LWR-type containment barrier including the first significant design for a closed cycle gas was not needed for future HTGRs 7). turbine plant, these designs did not materialize into the commissioning of a nuclear plant. This was primarily (3) Peach Bottom (No. 1) attributed to the nuclear industry being in a general state of decline in those countries with national HTGR Peach Bottom Unit 1 was the first prototype development programs. HTGR in the U.S. The 40MW(e) plant, owned and operated by the Philadelphia Electric Company, was built Performance of the THTR-300 and FSV in as part of the U.S. Atomic Energy Commission Power validating reactor safety characteristics and the TRISO Reactor Demonstration Program and the High coated fuel particle was very good.
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