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. However, FSV was Temperature Reactor Development Associates Program to beset with inconsistent operation and a corresponding low demonstrate the feasibility of a high performance, helium capacity factor throughout its life. The THTR-300 also had cooled, nuclear power plant. problems primarily associated with changing regulatory requirements that contributed to delays that eventually The reactor achieved initial criticality on March 3, ended up with a construction period of ~14 years. 1966, and the plant went into commercial operation in Corresponding financial concerns by the operating utilities June 1967. It operated successfully with a gross capacity led to the premature shutdown of both these plants in the factor of 74% and an overall availability of 88% (exclusive 1988-1989 time frame. of planned shutdowns for R&D programs) from June 1967 until October 31, 1974, when it was shut down for (1) Fort St. Vrain (FSV) decommissioning. Accumulated operation totaled 1,349 effective full-power days, for a total of 1,385,919 gross electrical megawatt hours generated. The FSV HTGR was operated by Public Service Company of Colorado as part of the U.S. Atomic Energy

Commission’s Advanced Reactor Demonstration Program. Peach Bottom provided a valuable demonstration This plant was designed with several advanced features of the high-temperature reactor concept by confirming the including; a.) A PCRV containing the entire primary core physics calculations, verifying the design analysis coolant system, b.) A core of hexagonal, graphite block methods, and providing a data base for further design fuel elements and reflectors with fuel in the form of activities in the following areas: reactor core; mechanisms ceramic coated (TRISO) particles, c.) Once-through steam in high-temperature helium; helium purification systems; 6) generator modules producing 538°C superheated main and and steam generator tube materials . reheat steam, and d.) Steam-turbine-driven axial helium circulators. At its rated capacity, the FSV plant generated 3. HTGR Demonstration Plants 842MW(t) to achieve a net output of 330MW(e).

Early international development of the HTGR The steam cycle was essentially conventional, focused on two basic core designs recognized specifically utilizing a standard reheat turbine, except that the steam by their fuel element structure. The German core design flow was from the high-pressure turbine exhaust to the has generally followed a core design incorporating helium circulator turbines before being reheated and spherical fuel elements, whereas, beginning with Fort St. returned to the intermediate pressure turbine. The steam Vrain (FSV), the U.S. core design included ceramic coated conditions were comparable to those of modern fossil fired fuel particles imbedded within rods placed in large power plants. hexagonal shaped graphite elements. The other major HTGR designer during this period was Russia with their Initial electric power generation was achieved at VG series of plant designs which incorporated the pebble FSV in December 1976, and 70% power was reached in bed core. November 1977. Full-power operation was achieved in 6) November 1981 . Although ~5.5 billion kwhr of The HTGR plants that followed the successful electricity was generated at FSV, the plant operated at low AVR and Peach Bottom included construction of FSV and availability primarily because of excessive downtime due the Thorium High Temperature Reactor (THTR-300). A to problems with the water-lubricated bearings of the major shift occurred with these plants including primary helium circulators. In spite of this low availability, the systems enclosed within PCRVs rather then steel vessels, plant was a valuable technology test-bed, successfully and accompanied by significant increases in plant power demonstrating the performance of several major systems,

including the reactor core with TRISO coated fuel particles the 1973 oil crisis and corresponding economic setback in hexagonal graphite blocks, reactor internals, steam and collapse of the U.S. nuclear power market of 1975. generators, fuel handling and helium purification 3). These plants incorporated cores of hexagonal graphite blocks with TRISO coated fuel particles similar to FSV. (2) Thorium High Temperature Reactor (THTR-300) The German HTR-500 made considerable use of The THTR-300 nuclear power plant included a the technology development for the THTR-300, with simplifications and optimizations based on practical steam cycle for the generation of electric power with a net 4) output of 296MWe. The heat generated in the reactor core experiences with the THTR-300 . This plant featured a of the helium cooled, graphite moderated high temperature simple design with the primary system components located reactor was transported via the helium gas coolant circuit within a single cavity PCRV, and included a pebble bed (primary system) to the steam generators where it was reactor power level of 1,390MWt to produce 550MWe of transferred to the steam-feedwater circuit (secondary electricity. system) and then transported to the turbine generator. The secondary system was cooled by means of a 180 m high The Russian HTGR development program natural draught dry cooling tower 4). included the VG series of plants primarily developed at OKBM in Nizhny Novgorod. Of these, the VG-400 design incorporated a pebble bed reactor with a final power level The THTR-300 power plant was sponsored by the of 1,060 MWt for co-generation applications of electricity Federal Republic of Germany and Nordrhein Westfalen generation and process heat production for steam (NRW). Construction of this 300MWe plant began in 1971, reforming of methane. The reactor was designed for a core but primarily due to increasing licensing requirements, the helium outlet temperature of 950°C. During the plant was not completed until 1984. This pebble bed preliminary phases of the design development both the reactor plant was connected to the electrical grid of the pebble bed and prismatic fuel block variants of the core utility Hochtemperatur-Kernkraftwerk GmbH (HKG) in were analyzed. As a result of the design and engineering November 1985. analysis, the pebble bed core was chosen for further development due to considerations of simplified fuel In August 1989, the decision was made for the element manufacturing technology and possibility of full permanent shutdown of the THTR-300. This action was scale testing in experimental reactors, utilization of a not due to technical difficulties associated with the plant, simplified core refueling mechanism and the possibility of but was the result of an application by HKG for early core refueling during on-load reactor operation. Also decommissioning based on a projected short fall in funding considered was a gas turbine cycle/VG-400 reactor plant. and contractual changes in the allocation of decommissioning costs between the FRG, NRW and HKG (1) Early Gas Turbine HTGR Assessment that would take effect upon the termination of the demonstration phase in 1991. A promising approach for making good use of the high temperature capability of the HTGR is to use the Operation of the THTR-300 was successful in primary helium coolant to drive a gas turbine in a direct validating the safety characteristics and control response of closed cycle arrangement. In the 1970s, this was the pebble bed reactor, primary system thermodynamics extensively studied in the U.S., Germany, the U.K. and and the good fission product retention of the fuel elements France. At that time, the concept was based on enclosing a 7). The plant operated over 16,000 hrs. and generated 2.891 8) large (2,000 to 3,000MWt) reactor core and the gas turbine billion kwhr of electricity . power conversion system within a prestressed concrete reactor vessel 24). One of the early gas turbine designs was 4. Large HTGR Steam Cycle Plant Designs 7) the HTGR-GT by General Atomic Co. This plant was projected to have the potential for high plant efficiencies The primary focus of international HTGR (40%) under dry cooling conditions and achievement of designers in the 1970s and early 1980s was associated with even higher efficiencies (50%) when bottoming cycles the development of large HTGR units. Continued interest were incorporated. An assessment of this plant was in development of larger steam cycle HTGR plants subsequently conducted in ~ 1980 with the focal point for included the German HTR-500, the Russian VG-400 and the assessment being the potential commercial HTGR-GT the United States’ HTGR-Steam Cycle (SC) plants. plant of 2000 MW(t)/800MW(e) and an average core outlet temperature of 850°C. In the U.S., the focus in the early 1970s was on HTGR-Steam Cycle designs of 770 to 1,160MWe. The principal findings of the assessment (in the Contracts to General Atomic Co. from U.S. utilities early 1980s) were as follows: (1) the HTGR-GT is feasible, included 10 plants that did not materialize due primarily to but with significantly greater development risk than that

associated with the HTGR-SC; (2) at the level of performance corresponding to the reference design, no incremental economic incentive can be identified for the HTGR-GT to offset the increased development costs and risk relative to those for the HTGR-SC (this was true over the range of cooling options investigated); (3) the relative economics of the HTGR-GT and HTGR-SC are not significantly affected by dry cooling considerations; and (4) although the reduced complexity of the cycle may ultimately result in a reliability advantage for the HTGR- GT, the value of that potential advantage could not be quantified.

Although these (early 1980s) findings did not provide the basis for the HTGR-GT as the preferred lead commercial plant, the HTGR-GT continued to engender Fig. 1 Maximum accident core temperature depicted interest from participating utilities. This interest stemmed chronologically for U.S. HTGR designs 3) from the potential of the HTGR-GT for improved efficiencies at high core outlet temperatures, the basic (1) HTR-MODULE 4) simplicity of the gas turbine cycle (low maintenance, high capacity factors), low water use requirements and 6) The 80MW(e) HTR module (HTR-MODULE) attractive co-generation characteristics . concept, developed by Siemens/Interatom, was the first small, modular type HTGR concept to be proposed (Fig. 2). In retrospect, this plant, although significantly Although initially developed in the early 1980s for different in general configuration compared to the current industrial process heat applications, the passive safety modular HTGR/gas turbine designs, its basic cycle has features of the side-by-side concept, coupled with the other remained unaltered throughout these past twenty years. attractive characteristics of the modular concept, soon led to the proposed electricity generation application. Work on a generic, site independent safety assessment of the HTR 5. Modular HTGR Steam Cycle Plant Development module was initiated with the filing, in 1987, of the safety analysis report in the State of Lower Saxony. The HTR The overall good safety characteristics of all module safety concept is characterized by comprehensive HTGRs are due to: the high heat capacity of the graphite protection of the environment by passive system core; the high temperature capability of the core characteristics even under extreme, hypothetical accident components; the chemical stability and inertness of the conditions. fuel, coolant, and moderator; the high retention of fission products by the fuel coatings; the single phase The safety features of the HTR module were characteristics of helium coolant; and the inherent negative based on the design condition that, even in the case of temperature coefficient of reactivity of the core. Fig. 1 failure of all active cooling systems and complete loss of provides a graphic overview of maximum accident core coolant, the fuel element temperatures would remain temperatures for U.S. HTGR designs. within limits such that there is virtually no release of radioactive fission products from the fuel elements. Such a The modular HTGR adds the unique condition guaranteed that the modular HTR power plant characteristic of being able to cool the reactor entirely by would not cause any hazard to the environment either passive heat transfer mechanisms following postulated during normal operation or in the case of an accident. accidents without exceeding the failure temperature of the coated particles. This characteristic has been achieved by A small core diameter (~3m) stems from the deliberately decreasing the core power level and requirement for reactor shutdown from all operating configuring the reactor so that natural heat removal conditions using only free falling control rods in reflector processes can limit fuel temperatures to levels at which the borings. The requirement to keep the maximum fuel release of fission products from the reactor system to the element temperature for all possible accidents inherently environment is insignificant for postulated accidents. Even below 1600°C, a temperature at which all radioactive for extreme accidents having very low probabilities of fission products are contained within the fuel elements, 3 occurrence, the cumulative fission product release at the leads directly to a mean power density of 3MW/m . In site boundary is estimated to be below those acceptable order to gain as much power as possible from the core, the under defined protective action guidelines 7). core height was chosen as large as possible.

availability and capacity factors and potential close-in siting required by industrial plants. The VGM modular HTGR, with a power output of 200MW(t) and a configuration very similar to the German HTR-MODULE was selected as a low power pilot unit after considering several different designs. The power required by large industrial complexes would be reached by the use of several such modules, providing the necessary reserves for both high availability and high capacity factors.

(4) MHTGR

In 1983, the U.S. organization representing utility/user interests in the HTGR program, the Gas Cooled Reactor Associates, conducted a survey to determine the utility nuclear generation preference for the future. This survey resulted in a strong interest for smaller generation increments. This was an important input leading to the evaluation and subsequent selection of the modular HTGR in 1985 3). Following a detailed evaluation in the spring and summer of 1985, a side-by-side concept similar in configuration to a German module design was chosen as the reference concept for further design and development by the U.S. program. The basic module was designed to deliver superheated steam at 17.3MPa and 538°C. An initial module power level of 250MWt was selected, but subsequent detailed safety analyses showed that this power Fig. 2 HTR-MODULE arrangement level could be increased with the hexagonal graphite block core design without compromising margins. Adopting an annular core allowed the power level to be increased (2) HTR-100 initially to 350MWt. Other reactor design changes and

analysis refinements subsequently allowed the power to be With the objective that nuclear power plants increased to 450MWt, while maintaining adequate margins utilizing small HTGRs can provide economic, 10) to component and safety limits . environmentally favorable and reliable electricity and heat for community and industrial purposes, Brown, Boveri und Central to the Modular HTGR (MHTGR) passive Cie and Hochtemperatur-Reacktorbau initiated the design safety approach was the annular reactor core of prismatic of the HTR-100 pebble bed plant. This design featured a fuel elements within a steel reactor vessel. A low-enriched 285MWt HTGR with a net electrical output of 100 MW on uranium, once-through fuel cycle was used. For a standard the basis of the AVR concept and utilized the advanced steam cycle MHTGR plant, the steam output from each of technologies of THTR-300 components and systems. The the four modules was connected to an individual turbine primary system included the reactor, steam generator and generator. The four module plant consists of two separate helium circulator in a single, vertical steel pressure vessel. areas, the nuclear island and the energy conversion area. The reactor core included 317,500 spherical elements of

TRISO type particles and a power density of 4.2 MW/m3. The most fundamental characteristic of the In the equilibrium cycle, 55% of the spherical elements MHTGR that separated it from previous reactor designs were fuel, with the remainder being graphite. The basic was the unique safety philosophy embodied in the design design for this plant incorporated two HTR-100 units with 11) . First, the philosophy requires that control of overall capability of producing 170 to 500 t/hr of industrial radionuclides be accomplished with minimal reliance on steam at 270°C and 16 bar, with 100 to 175MW(e) gross active systems or operator actions; the approach to safety output 9). is to rely primarily on the natural processes of thermal

radiation, conduction, and convection and on the integrity (3) VGM of the passive design features. Arguments need not center on an assessment of the reliability of pumps, valves, and Analysis of the heat energy market in Russia their associated services or on the probability of an revealed a need for the development of small nuclear operator taking various actions, given the associated power energy sources whose power utilization factor and uncertainties involved in such assessments. high degree of safety could meet the requirements of high

Second, the philosophy requires control of It was the development of the above mentioned releases by the retention of radionuclides primarily within modular HTGR steam plants that provided the key the coated fuel particle rather than reliance on secondary emphasis to initiation of the HTGR coupled to a gas barriers (such as the primary coolant boundary or the turbine power conversion system. With few exceptions, it reactor building). Thus, ensuring that the safety criteria are is the German HTR-MODUL and the HTR-100 reactor met is the same as ensuring that the retention capability of designs that are being utilized by ESKOM as the base for the coated fuel particles is not compromised. the PBMR. The MHTGR forms the basis for the GT-MHR reactor design 7). Both PBMR and GT-MHR reactors utilize an annular core arrangement.

III. Current and Projected HTGR Development Activities 2)

It was during the 1980s and early 1990s that a major focus of the international HTGR community was directed towards investigation of the industrial heat applications and co-generation capabilities provided by the modular HTGR. Investigation into the capabilities of this nuclear energy source to supply high temperature process heat as well as the need to evaluate the safety and technological features of the HTGR were contributing factors in the decisions by the Japan Atomic Energy Research Institute (JAERI) and China’s Institute of Nuclear Energy Technology (INET) to construct the High Temperature Engineering Test Reactor (HTTR) and High Temperature Gas Cooled Reactor Test Module (HTR-10), respectively.

It was also in the early 1990s that the emphasis of HTGR development moved from a balance of plant that incorporated the Rankine Steam Cycle to the gas turbine (Brayton Cycle) plant for the generation of electricity. This

3) major re-direction in research and development is the Fig. 3 MHTGR arrangement outgrowth of the notable improvement in net cycle

efficiency through the use of the Brayton Cycle coupled The assessment of the capability of the MHTGR with the corresponding projected decrease in cost brought to control accidental radioactivity releases shows that the about by plant simplification and modularization of the doses are a small fraction of the U.S.10CFR100 HTGR. requirements even for the bounding analyses which consider only the systems, structures and components that The current and projected development activities require neither operator action nor other than battery of research, design and testing the next generation of power. In fact, the exposures are so low that the protective HTGR plants is the subject of the following sections. It is action guidelines would require no evacuation or sheltering anticipated that the focus of nuclear power development plans for the public as specified in the utility/user will include a much greater emphasis on use of this energy requirements. The evaluation confirms that accident dose source for process heat applications and not continue the criteria can be met with a containment system that places present, almost exclusive, role of electricity production. primary emphasis on fission product retention within the fuel barriers. 1. HTGR Powered Process Heat Applications

Consistent between the modular HTGR designs Nuclear energy is now being used to produce of the HTR-MODULE, MHTGR and the VGM are side- about 17% of the world’s electricity. As of 1998, this by-side steel vessels housing the reactor and the helium included 442 nuclear reactors, with a total capacity of circulator and steam generator in a common configuration. about 354 gigawatts-electric (GWe). Yet, only a few of Fig. 3 depicts the U.S. MHTGR. The VGM is similar with these plants are being used to supply hot water and steam. the inclusion of another vessel housing an auxiliary The total capacity of these plants is about 5 GWt, and they cooling system. are operating in just a few countries, mostly in Canada,

China, Kazakhstan, Russia and the Ukraine. Water cooled be connected independently to a steam turbine and/or other reactors offer heat up to 300°C. These reactor types steam utilizing systems 13). include PWRs, BWRs, pressurized heavy water reactors, and light water cooled graphite moderated reactors. The production of hydrogen and subsequent Organic cooled heavy water moderated reactors reach synthesis to alternative, more environmentally acceptable temperatures of about 400°C, while liquid metal fast fuels such as methanol, is the primary focus of current breeder reactors produce heat up to 540°C. Gas cooled research into the process heat applications that would reactors reach even higher temperatures, about 650°C for effectively utilize the high temperature capabilities of the the AGR, and 950°C for the HTGR. HTGR.

For heat applications, specific temperature Many international organizations are focusing on requirements vary greatly. They range from low the modular HTGR as the energy source for the co- temperatures (~ room temperature), for applications such generation applications of high temperature industrial as hot water and steam for agro-industry, district heating, processes and electricity production. In addition to and seawater desalination, to ~ 1000°C for process steam hydrogen production, processes such as heavy oil recovery, and heat for the chemical industry and high pressure coal liquification and gasification are being explored. The injection steam for enhanced oil recovery, oil refinery capability for high temperature process heat applications is processes, and refinement of coal and lignite. The process key to the selection of the GFR and VHTR as Generation of water splitting for the production of hydrogen is at the IV advanced reactor designs (see section 4). upper end. Up to about 550°C, the heat can be supplied by steam; above that, requirements must be served directly by process heat due to the need for high steam pressures. An 2. High Temperature Test Reactors upper limit of 1,000°C for nuclear supplied process heat is set on the basis of the long-term strength capabilities of Interest in the HTGR as an advanced nuclear metallic reactor materials 12). power source for co-generation applications of electricity production and high temperature heat for industrial It is this unique capability of the HTGR as a processes has resulted in the construction of the HTTR by nuclear energy source to provide the broad range of JAERI and the HTR-10 by INET of Tsinghua University temperatures required to support a myriad of industrial in Beijing, China. These nuclear test facilities have the heat applications that is the emphasis for scientific capability of achieving core outlet temperatures to 950°C investigation and development currently being pursued on and 900°C, respectively, and will be utilized to support an international basis throughout the HTGR community. research and development activities including validation of An example of the use of the modular HTGR for industrial HTGR safety and general performance characteristics, applications is the MHTGR by General Atomics (GA). electricity generation via the gas turbine and validation of high temperature process heat applications. Key design GA is investigating the capabilities and flexibility characteristics of the HTTR and HTR-10 test reactor plants of the MHTGR to provide superheated steam for a are located herein in Appendix A. multitude of industrial applications. Energy requirements of industrial process complexes vary widely, according to (1) HTTR varying steam conditions, capacity requirements, and the ratio of thermal to electric power. The high temperature Construction of the HTTR at JAERI’s research and pressure steam at 17.3MPa and 540°C produced by the facilities in Oarai, Japan, began in 1991. First criticality Process Steam/Co-generation Modular Helium Reactor was obtained November 1998 with full power (at an (PS/C-MHR) can provide energy for heat cycles in a wide average core outlet temperature of 850°C) achieved in range of process applications and industrial complex sizes December 2001. This plant includes a 30MWth graphite- and capacities. The PS/C-MHR is being designed to meet moderated, helium cooled reactor comprised of hexagonal the rigorous requirements established by the U.S. NRC and fuel elements with TRISO coated fuel particles in compact the electric utility-user industry for a second generation form. Cooling of the HTTR consists of a main cooling power source for the future. system, an auxiliary cooling system and a vessel cooling system. The main cooling system is composed of a The most economic PS/C-MHR plant primary cooling system, a secondary helium gas cooling configuration includes an arrangement of several identical system and a pressurized water cooling system. The modular reactor units, each located in a single reactor primary cooling system has two heat exchangers, an 14) building. The plant is divided into two major areas: the intermediate heat exchanger and a primary water cooler . energy conversion area, containing turbine generators, a nuclear island containing the several reactor modules, and The HTTR will be utilized to establish and other balance of plant equipment. Each reactor module can improve on HTGR related technology, for the performance of innovative research, as a test facility for fuel and

materials irradiation, and to demonstrate process heat • The HTGR reactor size had been reduced in applications such as hydrogen production. Relative to developing the passively safe module design. At commercialization of the closed cycle gas turbine HTGR the same time, the size of industrial gas turbines plant, the HTTR will be a valuable facility in had increased. The technology was now available demonstrating the safety features of the HTGR, for a single turbo-machine to accommodate the performance and qualification of materials and heat energy from a single HTGR module. Highly components and in fuel and fission product behavior 15). effective compact recuperators had been developed. Recuperator size and capital (2) HTR-10 equipment cost are key economic considerations. • Highly effective plate-fin recuperators are much The HTR-10 Test Module is located at the INET smaller than equivalent tube and shell heat research site northwest of Beijing and is a major project in exchangers, provide for substantially less the Chinese National High Technology Program. complexity and capital cost, and are a key Construction of this plant was completed in the fall of requirement for achieving high plant efficiency. 2000. Initial criticality was achieved on 1 December with • The technology for large magnetic bearings had full power obtained in February 2003. been developed. The use of oil lubricated bearings for the turbo-machine with the reactor coolant The HTR-10 is a pebble bed 10MWth HTGR directly driving the turbine was problematic with with a primary helium circuit pressure of 3.0MPa. This test regard to the potential coolant contamination by reactor is the first step of the HTGR development strategy the oil. The availability of magnetic bearings in China. The objectives of the HTR-10 are to verify and eliminates this potential problem 19). demonstrate the technical and safety features of the modular HTGR and to establish an experimental base for A major requirement was for the plant to become developing nuclear process heat applications, including substantially simplified in order to provide a significant testing of electricity/heat co-generation and gas turbine reduction in the capital expenditure for new capacity 16) technology . additions. Figs. 4 and 5 provide a comparison of nuclear power plant efficiencies and the simplification that can be 3. Modular HTGR Gas Turbine Plant Development achieved in moving from the HTGR steam cycle to the basic direct gas turbine cycle, respectively. It has long been recognized that substantial gains in the generation of electricity from nuclear fission can be achieved through the direct coupling of a gas turbine to a HTGR. This advanced nuclear power plant is unique in its use of the Brayton cycle to obtain a net electrical efficiency approaching 50% 17). This plant provides a promising alternative for the utilization of nuclear energy to produce electricity. Although evaluation of this concept was initiated over twenty years ago, it was initially terminated due to the technical risks primarily in the component areas of magnetic bearing, compact plate-fin heat exchanger and turbo machinery development. Subsequent technological advancements in the design and operation of these components, coupled with the international capability for their fabrication and testing has Fig. 4 Plant efficiency comparison 3) resulted in renewed interest in this HTGR concept 18).

The possibilities presented by the gas turbine In order to be competitive, the thermal efficiency modular HTGR for substantial improvement in nuclear of nuclear power had to be markedly improved to compete power plant efficiency coupled with the potential for with modern, high efficiency fossil plants. HTGR significant gains in lowering capital and operating costs technology has always held the promise for electricity due to plant simplification have brought about an generation at high thermal efficiency by means of a direct increasing interest by international research organizations Brayton cycle and fortuitously, technological and plant developers. Overviews of the PBMR and GT- developments during the past decade provided the key MHR are provided in the following sub-sections. elements to realize this promise. These key elements are as Appendix A contains plant characteristics of many gas follows: turbine plant designs under investigation by Member States of the IAEA.

recuperator, precooler, intercooler and turbomachine as shown in Fig. 6.

(a) Steam cycle

(b) Closed cycle

Fig. 5 System simplification, steam cycle to the closed cycle gas turbine plant 3)

(1) Gas Turbine-Modular Helium Reactor (GT-MHR)

Renewed interest in the nuclear powered closed cycle gas turbine system within the U.S. resulted in the 25) present GT-MHR developmental program beginning in Fig. 6 GT-MHR PCS arrangement early 1993. Subsequent discussions with organizations of the Russian Federation resulted in GA and MINATOM Work performed in the preliminary design phase entering into a Memorandum of Understanding to has included full scale tests of a recuperator heat exchange cooperate on the development of the GT-MHR with the element, development of lab-scale technologies for goal, following design and development, to construct, test production of coated plutonium fuel and testing of and operate a prototype in Russia 22). FRAMATOME materials for the high temperature elements of the joined this program in 1996 with Fuji Electric becoming a turbomachine. Work continues in 2003 on the detailed participant and sponsor in 1997. The US government also design of selected systems and components including provides limited financial support on a matching resource initiation of the reactor system and power conversion unit basis with MINATOM for the destruction of weapons development tests, fabrication of initial test fuel and plutonium. Most of the design work on the GT-MHR is carbon-carbon composites, initiation of pre-licensing being performed within the nuclear organizations of the activities with the Russian licensing authorities and Russian Federation with financial and development of the safety analysis and environmental management/technical support from all members of the impact reports. The schedule for the GT-MHR includes 23) consortium. commissioning of the first module in 2010 .

The conceptual design is complete on this Although the GT-MHR is initially to utilize a 600MWt/293MWe plant. The basic primary system of this plutonium fuel cycle which has the capability of achieving closed cycle gas turbine plant includes two side-by-side a burn-up approaching 95%, the versatility and flexibility steel vessels with an outside appearance similar to the of this core will allow for the application of a wide range MHTGR (Fig. 3). The reactor vessel houses a 600MWt of diverse fuel cycles. Fuel derived from uranium, thorium annular core comprised of hexagonal fuel elements. The and a variety of plutonium grades is under consideration power conversion system (PCS) vessel houses the for long term applications in the GT-MHR 15).

pressure turbo/compressor unit and intercooler; the low (2) Pebble Bed Modular Reactor (PBMR) pressure turbo/compressor unit and precooler; and the power turbine/generator unit and recuperator, respectively. Table 1 provides key design characteristics for the The PBMR was first identified by ESKOM, the demonstration PBMR plant. electric utility of South Africa, in 1993 as an option for expansion of their electrical generation capacity. Subsequently, ESKOM contracted with Integrators of System Technology to perform a technical and economic study of the feasibility of the PBMR for the generation of electricity. This study, which was completed in early 1997, supported the continued development of the PBMR. Reactor development follows the HTR-MODULE pebble bed which was previously licensed in Germany for commercial operation. However, the PBMR utilizes an annular core with a fixed central column.

A series of internal reviews by ESKOM and subsequent independent reviews of this feasibility study by international entities have generally acknowledged that the PBMR is technically and economically capable of meeting the requirements originally set by ESKOM for commercialization. These requirements included: a.) New generation capacity capable of being located where the load growth is taking place in South Africa (coastal regions); b.) Small, modular increments of electrical generation capacity corresponding to system growth needs; c.) Reduced exposure to negative environmental issues such as carbon dioxide emissions and capable of providing Fig. 7 PBMR primary cycle a strategy for economic mitigation of greenhouse gas reductions; d.) Generating plants placed where there would Table 1 PBMR demonstration plant characteristics 26) be a limited need for extensive transmission system Plant Parameter Value additions, and; e.) Cost of capital and plant operation to be Reactor Output 400 MWt within those costs presently being achieved at ESKOM's Net Electrical Output ~ 170 MWe largest coal fired stations (with targets of capital cost less Cycle Efficiency 41.6% avg. than US$ 1,000/installed kw and overall generation, Core Inlet Temperature 489°C transmission and distribution costs of utility operations Core Outlet Temperature 900°C 15) within 2.0 cents/kwhr ) (Note: as of May 2003, the Helium Pressure 9MPa at 100% power 29) overnight cost requirement is < $1500/kW installed ). Avg. Core Power Density 4.78 MW/m3 High and Low Pressure 91% Over the past five years ESKOM, British Nuclear Turbine Efficiency Fuel and the Industrial Development Corporation of South High and Low Pressure 90.5% Africa have been developing the PBMR design and Compressor Efficiency performing detailed assessments for the introduction of Power Turbine Generator 92.5% this advanced nuclear power plant as additional generation Efficiency capacity on the ESKOM electric grid and for commercialization in the international market place 7). 4. International Research and Development Activities On 16 May 2003 the decision was made to proceed with the final design development and The technological and economic possibilities construction of a demonstration unit at the ESKOM presented by the modular HTGR gas turbine plant for the Koeberg nuclear plant site north of Cape Town. This generation of electricity as well as high temperature demonstration plant features a pebble bed annular core process heat for applications such as hydrogen production reactor housed within a steel vessel connected to a direct are currently under investigation by many Member States cycle gas turbine power conversion system. Fig. 7 is a of the IAEA. A tabulation of plant characteristics for these depiction of the PBMR primary cycle. The power plants is provided in Appendix A. conversion system includes three major steel vessels with associated interconnecting piping housing the high

Examples of these diverse plant designs include a.) Topics within the area of reactor physics and fuel the GTHTR 300 being developed by JAERI and the cycle studies (HTR-N) include; ACACIA plant in the Netherlands. • the qualification of tools and methodologies for HTR applications, The Japanese GTHTR 300 plant incorporates a • the analysis of the physics of HTR cores and 600MWt hexagonal fuel block core. The power conversion of possible HTR fuel cycles, mostly focused system includes a vertical heat exchanger vessel and a on the feasibility of cores fed only with civil horizontal turbo-machine vessel to allow for bearing plutonium from the 1st or 2nd generation, support and stable rotor operation. The cycle configuration • the comparison of the wastes from LWRs and is considerably simplified in comparison to the GT-MHR HTRs, and PBMR by elimination of compressor unit(s) and • the behaviour of irradiated HTR fuel in the corresponding intercooler. The overall net efficiency for conditions of deep geological disposal. this electricity producing plant is projected at ~45%. b.) Within the area of fuel technology, the HTR-F The ACACIA plant is being designed by NRG as objectives have included; a small combined heat and electricity producer for industry. • to restore the state-of-the-art technology on This plant incorporates a 40MWt pebble bed HTGR to HTR fuel fabrication, produce 14MW of electricity and 17 tonnes of 10 bar, • to develop a competitive HTR fuel, that is 220°C steam per hour. with very good performances in terms of high

21) burnup and temperature, with very low (1) HTR Technology Network (HTR-TN) Programme fission product releases and with a competitive fuel cycle cost, In Europe about 20 industrial and research • to verify the past tests on the ability of HTR organisations created the HTR Technology Network fuel to reach extremely high burnups (HTR-TN), to pool their effort of development of the (~ 700 GWd/tHM) needed to exploit the technologies needed for a first industrial deployment of potential of HTRs for burning plutonium modular direct cycle HTRs in the next decade and for efficiently. longer term improvements of this type of reactor. Significant support has been received from the European c.) The objective of the material development (HTR- Commission in its 5th Euratom Framework Programme M) project is to qualify the key materials for HTR (FP5) for the development of HTR technology in the fields applications: the reactor vessel material, the high of reactor physics, fuel and fuel cycle, materials, temperature (~ 850°C) materials for the turbine and components and licensing issues. In this context an reactor internal structures (e.g., control rod cladding) important programme of irradiation of HTR fuel, vessel and the graphite. The graphite grades used in the material and graphite is under preparation, and the former HTR projects are no longer available and the technology of fabrication of coated particle fuel is being suitability for HTR applications of the present mastered. International co-operation is also being industrial products is not known. Therefore a developed with the main organisations active in HTR programme of selection of appropriate grades was development outside of Europe. undertaken and in particular an irradiation of samples nd from different manufacturers started in the 2 half of The objectives shared by HTR-TN partners are: 2002 in HFR. For the vessel, two design options exist: • to create the technological basis for an industrial emergence of HTR in Europe by 2010, • the cooled vessel (PBMR), for which existing • to incite long term evolutions (design PWR vessel materials can be used without optimisation, temperature increase, opening of R&D, new application areas...) for improved • the hot vessel (GT-MHR), kept at the coolant competitiveness and fullfilment of the goals of inlet temperature (~450°C). Modified sustainable development. 9Cr1Mo steel has been selected for this

option. This material has many records of The following areas of research and development industrial applications but few nuclear ones, have been selected by the HTR-TN partners with a budget in particular in large thickness. of ~ 17 M€, of which 50% is funded by the EC:

d.) The objective of the HTR-E project is to assess the feasibility issues related to the components and technologies of the power conversion system (PCS) and to make recommendations for further developments. The key issues are identified from the

existing industrial projects (GT-MHR and PBMR). These issues include: (2) Generation IV Nuclear Power Technology

• The feasibility of a 300 MWe high-efficiency Ten countries have been involved in an helium turbine operating at 850°C. A preliminary international cooperative effort to assure that advanced design will be made and a programme of tests on nuclear energy systems are available for worldwide a mock-up for the validation of the performances deployment by 2030. These countries have agreed on a of this design will be defined. The realisation of framework of international cooperation in research known these tests will be proposed later for the follow-up as Generation IV. of the project in the 6th Framework Programme (FP6). Through the Generation IV International Forum • The recuperator, which is a key component for (GIF) these participating countries are defining the the thermal efficiency of the PCS, with heavy framework of research and development that will bring operating loads (temperature transients and about a future generation of nuclear energy systems that differential pressure) and stringent compactness can be licensed, constructed, and operated in a manner that requirements. A concept will be selected and a will provide competitively priced and reliable energy mock-up designed, fabricated and tested in CEA products while satisfactorily addressing nuclear safety, Grenoble, to validate the performances of the waste, proliferation and public perception concerns. The design in steady and transient conditions. mission of these energy systems will be electricity • Active magnetic bearings (and the necessary generation, hydrogen and process heat production, and 20) related catcher bearings) are the solution proposed actinide management . for supporting the turbo-machine in order to avoid lubrication of bearings, with the related risks of The very-high-temperature reactor and the gas- lubricant ingress. This technology exists, but the cooled fast reactor have been identified by the GIF as two mechanical and thermal loads for the GT-MHR of the six systems comprising Generation IV. are above the levels met in present industrial applications. Therefore a prototype of magnetic (a) Very High Temperature Reactor (VHTR) and catcher bearing will be designed, fabricated and tested on a test bench of the University of Development of a VHTR to achieve economically Dresden. competitive hydrogen production is a high priority of the • A leak tight rotating seal would allow the ability U.S. Department of Energy’s (DOE) nuclear programme. to have classical bearings outside of the PCS A bill authorizing the research, design and construction of vessel for the vertical support of the turbo- an advanced nuclear plant (anticipated as a VHTR) with machine and therefore to reduce the loads on the the capability for co-generation applications of electricity magnetic and catcher bearings and to make their generation and hydrogen production is under consideration design less challenging. It would also allow the in the U.S. Congress. Table 2 provides the design capability to put the alternator outside the primary characteristics for the VHTR. circuit, facilitating its design and maintenance. 20) Different concepts are being assessed, but it Table 2 VHTR design parameters seems that no existing technology can provide an Reactor Parameters Reference Value acceptable leakage. The possible ways of Reactor power 600 MWth improvement will be explored and a solution will Coolant inlet/outlet 640°/1,000°C be developed. Tests will be prepared for FP6. Temperature Core inlet/outlet pressure Dependent on process e.) The objective of HTR-L is to establish a European Helium mass flow rate 320 kg/s safety approach for modular direct-cycle HTRs. A large Average power density 6-10 MWth/m3 consensus on an approach taking into account the specific Reference fuel compound ZrC-coated particles in safety features of modular HTRs and leading to a blocks, pins or pebbles simplification in the reactor design (e.g., suppression of Net plant efficiency >50% engineered safety systems, limitation of the number of components with a safety classification, reduction of the The VHTR system utilizes a graphite moderated, loads on the third barrier…) is very important for helium cooled reactor with a thermal neutron spectrum and consolidating the economic competitiveness of HTRs. a once-through uranium cycle. This plant is primarily Therefore the results of HTR-L will not only be presented aimed at high temperature process heat applications, such to European safety authorities, but HTR-TN will also take as coal gasification and thermochemical hydrogen initiatives to incite large international discussions on these production with excellent efficiency. The VHTR has a results, in order to broaden the consensus. coolant outlet temperature of 1,000°C and therefore can

supply a broad spectrum of high temperature, non-electric the GFR and VHTR through the viability and performance applications. The reference 600MWth VHTR dedicated to phases of development. Table 4 provides an overview of hydrogen production can yield >2,000,000 m3 of hydrogen these costs broken down into general categories. per day. Also, the VHTR has the capability to be an attractive heat source for co-generation applications in Table 4 R&D costs for the GFR and VHTR 20) large industrial complexes. Area of Development GFR R&D VHTR R&D Cost-$ Cost-$ The goal set for the VHTR is for a demonstration Fuels & Materials 300 M 170 M plant in 2015. The summary schedule to achieve this goal Reactor Systems 100 M 20 M includes: Balance of Plant 50 M 280 M • An accelerated preconceptual/conceptual design Safety 150 M 80 M • Two step 10CFR license with the construction Design & Evaluation 120 M 90 M license in 2010 Fuel Cycle 220 M 30 M • An estimated construction time frame of 5 years Total 940 M 670 M and $600 M with an NRC operating license in 2015 The development requirements of each Gen. IV • A 2007 start date for procurement of long lead plant vary considerably. An example of some of the R&D items needs for the VGTR include the following: • Cost estimates include an average of $40 M/year • Fuel including the following: in 2004-2011 for design work ($25 M in FY04 o demonstration of the TRISO fuel to and rising to $80 M in FY09) required quality level and performance o fabricate test particles and test in (b) Gas-cooled Fast Reactor (GFR) prototypic reactor conditions (and accident conditions) The GFR features a fast neutron spectrum and o extend to high burnup closed fuel cycle for efficient conversion of fertile uranium (>200,000MWD/Mtu) and management of actinides. The referenced reactor is a o For core outlet temperatures >900°C, 600 MWth/288 MWe, helium cooled system operating fuel temperatures >1,300°C may be with an outlet temperature of 850°C using a direct Brayton needed. Examine higher temperature cycle gas turbine for high thermal efficiency 20) (Table 3). coatings such as ZrC and repeat fuel This high helium coolant temperature makes it possible to testing as indicated above. generate electricity or for hydrogen production and other process heat applications, all with high conversion. • Develop, test and design new control rod materials 20) Table 3 GFR design parameters • Graphite fabrication and qualification Reactor Parameter Reference Value • Decommissioning issues such as graphite disposal Reactor power 600 MWth • Select/qualify materials for high temperature Net plant efficiency 48% applications in key components such as the (direct cycle helium) reactor vessel, hot gas ducts and components, Coolant inlet/outlet 490°C/850°C recuperator, intermediate heat exchanger, etc. temperature and pressure at 90 bar • Test for effects on materials due to impurities in 3 Average power density 100 MWth/m the primary coolant helium Referenced fuel compound UPuC/SiC (70/30%) • Define and demonstrate the hydrogen production With about 20% Pu content processes Volume fraction, 50/40/10% • Complete R&D on the Power Conversion System Fuel/Gas/SiC components including the turbocompressor, Conversion ratio Self-sufficient magnetic bearings, recuperator, precooler and Burnup, Damage 5% FIMA; 60 dpa intercooler, seals and couplings, etc.

(3) Additional U.S. HTGR R&D Programmes (c) R&D for Generation IV Gas Cooled Reactors Along with Generation IV work mentioned above, Significant R&D will be necessary to achieve the a primary focus within the DOE advanced gas reactor goal of bringing any Generation IV plant to (AGR) programme is the continued development and commercialization. The GIF, through the technical qualification of coated fuel particles. This fuel programme working group on gas-cooled reactors, has provided a includes supporting the near term deployment of an AGR general quantification of the costs associated with bringing for energy production in the U.S. by reducing market entry

risks associated with fuel production and qualification and IV. Conclusions to establish a bridge to VHTR concepts and more advanced fuels 27). Deployment of the HTGR as an advanced source of nuclear energy for the 21st century incorporates a strong The DOE AGR fuel programme requirements developmental base spanning the past four decades include: including Dragon, the AVR, Peach Bottom, Fort St. Vrain, • Manufacture of high quality coated fuel particles and the THTR power plants. This base exhibited an (LEU with a prismatic gas cooled reactor design, excellent foundation for demonstrating the safety and GT-MHR focus) environmental aspects of the HTGR, particularly the high • Complete the design and fabrication of reactor temperature performance of UO2 kernel TRISO coated fuel test rigs for irradiation testing of coated fuel particle. forms • Demonstrate fuel performance during normal and This developmental base has been recently accident conditions via irradiation, safety tests enhanced by the construction of the HTTR and the HTR- and PIE 10 test reactors in Japan and China, respectively. These • Improve predictive fuel performance/behaviour plants are currently undergoing testing to further and fission product transport models demonstrate the safety and environmental capabilities of the HTGR to average core outlet temperatures approaching Within EPRI, cost reduction has been adopted as the 950°C. These plants will also be used to investigate and theme for their R&D programme specific to future nuclear validate materials and components for future high plants. R&D projects have been initiated on a prioritized temperature process heat applications. cost reduction basis as funding increments are available. Of the categories for this programme, EPRI projects The major emphasis of HTGR development in the leading to new products such as the HTGR gas turbine past two decades has been with the small, modular HTGR reactor and advancing the technology of the potentially with the flexibility of being coupled to a gas turbine for the competitive HTGR alternatives to the LWR are being generation of electricity or to other process outputs such as investigated. Projects that have been initiated include: steam for a multitude of high temperature industrial heat • Inspectability and maintainability (I&M): This applications. The current focus on the modular HTGR project applied EPRI’s LWR experience to an with an annular core is expected to contributed evaluation of the I&M associated with the HTGR substantially to the goals of reduced capital and operating gas turbine PCS. Detailed recommendations were costs by plant simplification and a significant improvement provided to the GT-MHR and PBMR projects in in cycle efficiency over existing nuclear plant designs. 2001 and 2002, respectively • Graphite endurance: This project details the It is significant to note that the constituency in the challenges of graphite irradiation, manufacture development of the HTGR has changed substantially from capabilities, oxidation, availability, etc. which the single nationalism of the past (i.e., Germany, Russia or must be addressed for a 60 year lifetime. the US) to the international involvement of many nations working together in the investigative research required to • Silver retention: This continuing project addresses the challenge posed by radioactive bring this advanced nuclear energy source to fruition. The reasons for this are obvious; the HTGR is a design that silver deposits on the PCS components. shows promise for significant improvements in nuclear Development of a deposition/penetration resistant coating for the most vulnerable plant economics and attendant plant efficiency, with the long-term prospect of achieving substantial marketability equipment surfaces is currently underway. and associated financial reward. However, the task ahead • Catcher bearings: This on-going project is to commercialize this advanced nuclear plant must addressing the requirements and test facility to overcome significant hurdles including: A modular HTGR validate the bearings required in the event of has not been licensed or constructed, and the size and failure of the magnetic bearings in the PCS. environmental application of major components such as • Helium seals: This project has now identified the plate-fin recuperators and turbomachines utilizing preferred helium seal for the PCS and has magnetic bearings has yet to be achieved for the gas contracted for the laboratory testing of this seal • turbine plant. Hydrogen production: The HTGR has a strong potential to be a major energy source for However, successful commercial deployment of hydrogen separation. EPRI has initiated a project this plant has the potential of significantly advancing that investigates the economic potential of the 28) nuclear power as a world-wide primary energy source for HTGR in this role . the future with corresponding financial reward to the

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Appendix A: Selected Characteristics of HTGR Plants

Steam Cycle HTGRs AVR Peach Ft. St. THTR- HTR-500 VGM- HTR- MHTGR Bottom Vrain 300 400 Module Country/Origin Germany U.S. U.S. Germany Germany Russia Germany U.S. Thermal Pwr, MWt 46 115 842 750 1,390 1,060 200 350 Net Elec. Pwr, MWe 13 40 330 300 550 Co-Gen. 80 139 Pwr Density, 2.5 8.3 6.3 6.0 6.6 6.9 3.0 5.9 MW/m3 Core Out Temp, °C 950 725 775 750 700 950 700 686 Helium Pres, MPa 1.1 2.25 4.8 3.9 5.5 5.0 6.6 6.4 Steam Temp. °C 505 538 538/538 530/530 530 535 530 538 Elec. Gen., MWh ~ 1,670 ~ 1,380 ~ 5,500 ~ 2,890 N/A N/A N/A N/A Reactor Type Pebble Sleeve Block Pebble Pebble Pebble Pebble Block Fuel Enrichment Various HEU HEU HEU LEU LEU LEU LEU Fuel Composition Oxide Carbide Carbide Oxide Oxide Oxide Oxide Ox-Carb Fuel Coating Various BISO TRISO BISO TRISO TRISO TRISO TRISO Vessel Material Steel Steel PCRV PCRV PCRV PCRV Steel Steel

HTGR Test Reactors HTTR HTR-10: GT-ST Configuration Country/Origin Japan China Thermal Pwr, MWt 30 10 Elec. Pwr, MWe ------GT~2, ST~1.4 Cycle Eff., % N/A 34.4 Core Out Temp, °C 950 max. 900 Helium Pres, MPa 4.0 3.0 GT Cycle Type N/A In-Direct Core Type Block Pebble Fuel Enrichment LEU LEU Fuel Composition Oxide Oxide Fuel Coating TRISO TRISO Vessel Material Steel Steel

Modular HTGR Gas Turbine Plants GT-MHR PBMR- MHTGR- ACACIA GTHTR- 600MW- MPBR demo. IGT 300 HTGR-GT Country/Origin Russia/U.S. S. Africa China Netherlands Japan Japan U.S. Thermal Pwr, MWt 600 400 200 40 600 600 250 Net Ele Pwr, MWe 278 ~ 170 ~ 96 Co-Gen. 273 287 112 Pwr. Den, MW/m3 6.5 4.78 3.0 ------5.77 ------Core Out Temp, °C 850 900 900 800 850 850 850 Helium Pres, MPa 7.15 9.0 6.0 2.3 6.8 6.0 7.9 Cycle Type Direct Direct In-Direct Direct Direct Direct In-Direct Core Type Block Pebble Pebble Pebble Block Pin/Block Pebble Fuel Enrichment HE-Pu LEU LEU LEU LEU LEU LEU Fuel Composition PuO Oxide Oxide Oxide Oxide Oxide Oxide Fuel Coating TRISO TRISO TRISO TRISO TRISO TRISO TRISO Vessel Material Steel Steel Steel Steel Steel Steel Steel