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Sample records for high-temperature nuclear reactor

  1. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  2. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  3. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  4. Technological improvements to high temperature thermocouples for nuclear reactor applications

    International Nuclear Information System (INIS)

    Schley, R.; Leveque, J.P.

    1980-07-01

    The specific operating conditions of thermocouples in nuclear reactors have provided an incentive for further advances in high temperature thermocouple applications and performance. This work covers the manufacture and improvement of existing alloys, the technology of clad thermocouples, calibration drift during heat treatment, resistance to thermal shock and the compatibility of insulating materials with thermo-electric alloys. The results lead to specifying improved operating conditions for thermocouples in nuclear reactor media (pressurized water, sodium, uranium oxide) [fr

  5. Nuclear reactor application for high temperature power industrial processes

    International Nuclear Information System (INIS)

    Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

    1977-01-01

    This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

  6. Neutron analysis of the fuel of high temperature nuclear reactors

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2014-10-01

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  7. High temperature reactor and application to nuclear process heat

    Energy Technology Data Exchange (ETDEWEB)

    Schulten, R; Kugeler, K [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.)

    1976-01-01

    The principle of high temperature nuclear process heat is explained and the main applications (hydrogasification of coal, nuclear chemical heat pipe, direct reduction of iron ore, coal gasification by steam and water splitting) are described in more detail. The motivation for the introduction of nuclear process heat to the market, questions of cost, of raw material resources and environmental aspects are the next point of discussion. The new technological questions of the nuclear reactor and the status of development are described, especially information about the fuel elements, the hot gas ducts, the contamination and some design considerations are added. Furthermore the status of development of helium heated steam reformers, the main results of the work until now and the further activities in this field are explained.

  8. Very high temperature measurements: Applications to nuclear reactor safety tests

    International Nuclear Information System (INIS)

    Parga, Clemente-Jose

    2013-01-01

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  9. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  10. High temperature friction and seizure in gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Cousseran, P.; Febvre, A.; Martin, R.; Roche, R.

    1978-01-01

    One of the most delicate problems encountered in the gas cooled nuclear reactors is the friction without lubrication in a dry and hot (800 0 C /1472 0 F) helium atmosphere even at very small velocity. The research and development programs are described together with special tribometers that operate at mode than 1000 0 C (1832 0 F) in dry helium. The most interesting test conditions and results are given for gas nitrited steels and for strongly alloyed Ni-Cr steels coated with chromium carbide by plasma sprayed. The effects of parameters live velocity, travelled distance, contact pressure, roughness, temperature and prolonged stops under charge are described together with the effects of negative phenomena like attachment and chattering [fr

  11. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  12. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  13. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

    International Nuclear Information System (INIS)

    Gotschall, H.L.

    1994-01-01

    The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

  16. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  17. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  18. Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

    International Nuclear Information System (INIS)

    Hodzic, A.; Kugeler, K.

    1997-01-01

    This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

  19. Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

    2005-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various

  20. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

    International Nuclear Information System (INIS)

    Beech, D.J.; May, R.

    2000-01-01

    In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

  1. Hydrogen production by high temperature electrolysis of water vapour and nuclear reactors

    International Nuclear Information System (INIS)

    Jean-Pierre Py; Alain Capitaine

    2006-01-01

    This paper presents hydrogen production by a nuclear reactor (High Temperature Reactor, HTR or Pressurized Water Reactor, PWR) coupled to a High Temperature Electrolyser (HTE) plant. With respect to the coupling of a HTR with a HTE plant, EDF and AREVA NP had previously selected a combined cycle HTR scheme to convert the reactor heat into electricity. In that case, the steam required for the electrolyser plant is provided either directly from the steam turbine cycle or from a heat exchanger connected with such cycle. Hydrogen efficiency production is valued using high temperature electrolysis. Electrolysis production of hydrogen can be performed with significantly higher thermal efficiencies by operating in the steam phase than in the water phase. The electrolysis performance is assessed with solid oxide and solid proton electrolysis cells. The efficiency from the three operating conditions (endo-thermal, auto-thermal and thermo-neutral) of a high temperature electrolysis process is evaluated. The technical difficulties to use the gases enthalpy to heat the water are analyzed, taking into account efficiency and technological challenges. EDF and AREVA NP have performed an analysis to select an optimized process giving consideration to plant efficiency, plant operation, investment and production costs. The paper provides pathways and identifies R and D actions to reach hydrogen production costs competitive with those of other hydrogen production processes. (authors)

  2. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    International Nuclear Information System (INIS)

    Beck, J.M.; Collins, J.W.; Garcia, C.B.; Pincock, L.F.

    2010-01-01

    High Temperature Gas Reactors (HTGR) have been designed and operated throughout the world over the past five decades. These seven HTGRs are varied in size, outlet temperature, primary fluid, and purpose. However, there is much the Next Generation Nuclear Plant (NGNP) has learned and can learn from these experiences. This report captures these various experiences and documents the lessons learned according to the physical NGNP hardware (i.e., systems, subsystems, and components) affected thereby.

  3. Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

    International Nuclear Information System (INIS)

    Liu Yu; Dong Yujie

    2011-01-01

    In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

  4. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  5. Combined conditioning in the high-temperature experimental nuclear reactor (AVR) at Juelich

    International Nuclear Information System (INIS)

    Nieder, R.; Vey, K.; Ivens, G.

    1984-01-01

    The high temperature experimental nuclear reactor (AVR) is the first nuclear power plant in which combined cycle operation has been introduced. The water-steam cycle has been operated for about 15 years according to the alkali method of working with ammonia and hydrazine. The VGB-guidelines have been adhered to througout. Since January 1983 cobined cycle operation has been employed, and in this process a pH-value of about 8.5 and an oxygen concentration of about 200 μg/kg in the feedwater have been used. A distinct reduction of tritium concentration in the water-steam cycle was the outstanding new result. (orig.) [de

  6. Microstructural control and high temperature mechanical property of ferritic/martensitic steels for nuclear reactor application

    International Nuclear Information System (INIS)

    Adetunji, G.J.

    1991-04-01

    The materials under study are 9-12% Cr ferritic/martensitic steels, alternative candidate materials for application in core components of nuclear power reactors. This work involves (1) Investigation of high temperature fracture mechanism during slow tensile and limited creep testing at 600 o C (2) Extensive study of solute element segregation both theoretically and experimentally (3) Investigation of effects by thermal ageing and irradiation on microstructural developments in relation to high temperature mechanical behaviour. From (1) the results obtained indicate that the important microstructural characteristics controlling the fracture of 9-12% Cr ferritic/martensitic steels at high temperature are (a) solute segregation to inclusion-matrix interfaces (b) hardness of the martensitic matrix and (c) carbide particle size distribution. From (2) the results indicate a strong concentration gradient of silicon and molybdenum near lath packet boundaries for certain quenching rates from the austenitizing temperature. From (3) high temperature tensile data were obtained for irradiated samples with thermally aged ones as control. (author)

  7. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    International Nuclear Information System (INIS)

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-01-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

  8. Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

    International Nuclear Information System (INIS)

    Goto, Minoru

    2015-03-01

    An appropriate configuration of fuel and reactivity control equipment in a nuclear reactor core, which allows the design of the nuclear reactor core for low cost and high performance, is performed by nuclear design with high accuracy. The accuracy of nuclear design depends on a nuclear data library and a nuclear analysis method. Additionally, it is one of the most important issues for the nuclear design of a High Temperature Gas-cooled Reactor (HTGR) that an insertion depth of control rods into the reactor core should be retained shallow by reducing excess reactivity with a different method to keep fuel temperature below its limitation thorough a burn-up period. In this study, using experimental data of the High Temperature engineering Test Reactor (HTTR), which is a Japan's HTGR with 30 MW of thermal power, the following issues were investigated: applicability of nuclear data libraries to nuclear analysis for HTGRs; applicability of the improved nuclear analysis method for HTGRs; and effectiveness of a rod-type burnable poison on HTGR reactivity control. A nuclear design of a small-sized HTGR with 50 MW of thermal power (HTR50S) was performed using these results. In the nuclear design of the HTR50S, we challenged to decrease the kinds of the fuel enrichments and to increase the power density compared with the HTTR. As a result, the nuclear design was completed successfully by reducing the kinds of the fuel enrichment to only three from twelve of the HTTR and increasing the power density by 1.4 times as much as that of the HTTR. (author)

  9. HTGR nuclear power plants: features of the VGR-50 high temperature reactor

    International Nuclear Information System (INIS)

    Glebov, V.P.; Bogoyavlenskii, R.G.; Glushkov, E.S.; Grebennik, V.N.; Ponomarev-Stepnoi, N.N.; Vinogradov, V.P.

    1983-01-01

    Current developmental trends in the power industry are guided to an appreciable extent by the increasing shortages of fossil fuels (coal, petroleum, natural gas) and by ecological problems. Assuming a continuing trend in worldwide consumption of energy resources, we see the electric power industry using up 20%, the other 80% (petroleum, coal, natural gas) going into generating industrial process heat and space heat, transportation, the chemical processing industry, the metallurgical industry, and other branches of industry. In the future, nuclear power will have the job of not only meeting the needs of the electric power industry, but also generating process heat. The most promising type of nuclear power plant available for solving complex problems in generation of electric power and heat for technological processes in the metallurgical processing industry and chemical processing industry is the one based around high-temperature reactors

  10. Cost comparison of very high temperature nuclear reactors for process heat applications

    International Nuclear Information System (INIS)

    Crowley, J.H.; Newman, J.B.

    1975-03-01

    In April 1974, the United States Atomic Energy Commission (USAEC) authorized General Atomic Company, General Electric Company and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing very high temperature nuclear reactors. General Electric and Westinghouse produced concepts for the entire nuclear system, including the balance of plant. The General Atomic assessment included only the nuclear reactor portion of the nuclear plant. United Engineers and Constructors Inc. (UE and C) was requested by the USAEC in November 1974 to prepare an economic comparison of the three conceptual plants. The comparison is divided into three tasks: (1) Develop a balance of plant conceptual design to be combined with the General Atomic concept as a basis for comparison, and estimate the cost of the General Atomic/UE and C concept in July 1974 dollars; (2) Normalize the overall plant costs for the General Atomic/UE and C, General Electric and Westinghouse concepts, compare the costs, and identify significant differences between the concepts; and (3) Estimate the operation and maintenance costs for the General Atomic/UE and C plant and compare with the other concepts. The results of these task studies are discussed

  11. Coal gasification coal by steam using process heat from high-temperature nuclear reactors

    International Nuclear Information System (INIS)

    Heek, K.H. van; Juentgen, H.; Peters, W.

    1982-01-01

    This paper outlines the coal gasification process using a high-temperature nuclear reactor as a source of the process heat needed. Compared to conventional gasification processes coal is saved by 30-40%, coal-specific emissions are reduced and better economics of gas production are achieved. The introductory chapter deals with motives, aims and tasks of the development, followed by an explanation of the status of investigations, whereby especially the results of a semi-technical pilot plant operated by Bergbau-Forschung are given. Furthermore, construction details of a full-scale commercial gasifier are discussed, including the development of suitable alloys for the heat exchanger. Moreover problems of safety, licensing and economics of future plants have been investigated. (orig.) [de

  12. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  13. Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

    1987-01-01

    Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

  14. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  15. R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

    2005-01-01

    The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

  16. Nuclear power for coexistence with nature, high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

  17. Aspects of nuclear process heat application of very high temperature reactors (VHTR)

    International Nuclear Information System (INIS)

    Jansing, W.T.; Kugeler, K.

    2014-01-01

    The different processes of high temperature process application require new concepts for heat exchangers to carry out key process like steam reforming of light hydrocarbons, gasification of coal or biomass, or thermo-chemical cycles for hydrogen production. These components have been tested in the German projects for high temperature development. The intention was always to test at original conditions of temperatures, pressures and gas atmospheres. Furthermore the time of testing should be long as possible, to be able to carry out extrapolations to the real lifetime of components. Partly test times of around 20 000 hours have been reached. Key components, which are discussed in this paper, are: Intermediate heat exchangers to separate the primary reactor side and the secondary process side. Here two components with a power of 10 MW have been tested with the result, that all requirements of a nuclear component with larger power (125 MW) can be fulfilled. The max. primary helium temperature was 950°C, the maximal secondary temperature was 900°C. These were components with helical wounded tubes and U-tubes. In the test facility KVK, which had been built to carry out many special tests on components for helium cycles, furthermore hot gas ducts (with large dimensions), hot gas valves (with large dimensions), steam generators (10 MW), helium circulators, the helium gas purification and special measurements installations for helium cycle have been tested. All these tests delivered a broad know how for the urther development of technologies using helium as working fluid. The total test time of KVK was longer than 20 000 h. In a large test facility for steam reforming (EVAⅡ10 MW, T He =950°C, p He =40 bar, T Reform =800°C) all technical details of the conversion process have been investigated and today the technical feasibility of this process is valuated as given. Two reformer bundles, one with baffles and one with separate guiding tubes for each reformer tube have

  18. Development of helium turbines associated with high temperature nuclear reactors for electric power plants

    International Nuclear Information System (INIS)

    Chaboseau, J.

    1975-01-01

    First, the system is defined and found perfectly adapted, in the industrial meaning of the term, to the utilization with high temperature nuclear reactors. Subsequently, a realistic determination of its original feature and that of the turbine is tempted: these features are few, particularly if the existence of numerous, already utilized, direct cycle devices is considered. Then, the influence of main characteristic parameters is evaluated, i.e. the choice of thermodynamic cycle, the utilization of helium, the principle of the layout and the rating of the power station. It appears to be sure that the engineering designs are of great importance for the turbine designer. For the latter, however, the technical aspects of the required developments will be of the conventional type. The character and the importance of these developments will mainly depend on how the system is spread in industry. It seems possible that the reasonable choice of the data of the first generating station allows to reduce the preliminary developments to a minimum amount while ensuring the subsequent evolution of the techniques. The inherent research of perfectness of the system justifies to anticipate long-term developments which are pursued in parallel with the commissioning of subsequent generations of power stations [fr

  19. 9th international conference on high-temperature reactors - coal and nuclear energy for electricity and gas generation

    International Nuclear Information System (INIS)

    Kelber, G.

    1987-01-01

    The site of the high-temperatur reactor in the Ruhr region neighbouring on a coal-fired power plant is not accidental. The potential of the high-temperature reactor as a central plant element for the supply of heat for heating purposes and process heat covers also the possibility of coal gasification and liquefaction. Therefore the high-temperature reactor is, in the long term, a ray of hope for the coal region, able to compensate for the production-related competitive disadvantages of local coal. It can contribute to guaranteeing in the long term the task of German hard coal as an essential pillar of our energy supply. The VGB as a technical association of thermal power plant operators is particularly committed to the integration of coal and nuclear energy. Within the bounds of its possibilities, it will contribute to promoting the safe and environmentally beneficial generation of electricity from the two primary energy sources. (orig./DG) [de

  20. Use of a Nuclear High Temperature Gas Reactor in a Coal-To-Liquids Process

    International Nuclear Information System (INIS)

    Robert S. Cherry; Richard A. Wood

    2006-01-01

    AREVA's High Temperature Gas Reactor (HTGR) can potentially provide nuclear-generated, high-level heat to chemical process applications. The use of nuclear heat to help convert coal to liquid fuels is particularly attractive because of concerns about the future availability of petroleum for vehicle fuels. This report was commissioned to review the technical and economic aspects of how well this integration might actually work. The objective was to review coal liquefaction processes and propose one or more ways that nuclear process heat could be used to improve the overall process economics and performance. Shell's SCGP process was selected as the gasifier for the base case system. It operates in the range of 1250 to 1600 C to minimize the formation of tars, oil, and methane, while also maximizing the conversion of the coal's carbon to gas. Synthesis gas from this system is cooled, cleaned, reacted to produce the proper ratio of hydrogen to carbon monoxide and fed to a Fischer-Tropsch (FT) reaction and product upgrading system. The design coal-feed rate of 18,800 ton/day produces 26.000 barrels/day of FT products. Thermal energy at approximately 850 C from a HTGR does not directly integrate into this gasification process efficiently. However, it can be used to electrolyze water to make hydrogen and oxygen, both of which can be beneficially used in the gasification/FT process. These additions then allow carbon-containing streams of carbon dioxide and FT tail-gas to be recycled in the gasifier, greatly improving the overall carbon recovery and thereby producing more FT fuel for the same coal input. The final process configuration, scaled to make the same amount of product as the base case, requires only 5,800 ton/day of coal feed. Because it has a carbon utilization of 96.9%, the process produces almost no carbon dioxide byproduct Because the nuclear-assisted process requires six AREVA reactors to supply the heat, the capital cost is high. The conventional plant is

  1. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  2. Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

    International Nuclear Information System (INIS)

    Nieder, R.; Resch, G.

    1976-01-01

    The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

  3. Small high temperature gas-cooled reactors with innovative nuclear burning

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Ismail; Sekimoto, Hiroshi

    2008-01-01

    Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles. (author)

  4. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

  5. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

    2009-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

  6. Material development for gas-cooled high temperature reactors for the production of nuclear process heat

    International Nuclear Information System (INIS)

    Nickel, H.

    1977-04-01

    In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

  7. Pumps of molten metal based on magnetohydrodynamicprinciple for cooling high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Donátová, M.; Karban, P.; Ulrych, B.

    2009-01-01

    Roč. 85, č. 4 (2009), s. 13-15 ISSN 0033-2097 R&D Projects: GA ČR(CZ) GA102/07/0496 Institutional research plan: CEZ:AV0Z20570509 Keywords : pumps of molten metal * magnetohydrodynamic principle * nuclear reactors Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.196, year: 2009

  8. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  9. Modeling the high-temperature gas-cooled reactor process heat plant: a nuclear to chemical conversion process

    International Nuclear Information System (INIS)

    Pfremmer, R.D.; Openshaw, F.L.

    1982-05-01

    The high-temperature heat available from the High-Temperature Gas-Cooled Reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design

  10. Control room conceptual design of nuclear power plant with multiple modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Jia Qianqian; Qu Ronghong; Zhang Liangju

    2014-01-01

    A conceptual design of the control room layout for the nuclear power plant with multiple modular high temperature gas-cooled reactors has been developed. The modular high temperature gas-cooled reactors may need to be grouped to produce as much energy as a utility demands to realize the economic efficiency. There are many differences between the multi-modular plant and the current NPPs in the control room. These differences may include the staffing level, the human-machine interface design, the operation mode, etc. The potential challenges of the human factor engineering (HFE) in the control room of the multi-modular plant are analyzed, including the operation workload of the multi-modular tasks, how to help the crew to keep situation awareness of all modules, and how to support team work, the control of shared system between modules, etc. A concept design of control room for the multi-modular plant is presented based on the design aspect of HTR-PM (High temperature gas-cooled reactor pebble bed module). HFE issues are considered in the conceptual design of control room for the multi-modular plant and some design strategies are presented. As a novel conceptual design, verifications and validations are needed, and focus of further work is sketch out. (author)

  11. Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

    Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

  12. Plutonium burning in a pebble-bed type high temperature nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bende, E.E

    2000-01-24

    This thesis deals with the pebble-bed High Temperature Reactor that is fuelled with pure reactor-grade plutonium. It is stressed that neither burnable poisons nor fertile materials like 238U and 212Th are present in the calculational models throughout this thesis. Chapter 2 discusses the general properties of the pebble-bed HTR: the passive safety features of this reactor; different fuel scenarios according to which the pebble-bed HTR can be operated; properties of the pebbles and the coated particles (CPs), including a concise overview of the mechanisms that can lead to coated particle failure. Special attention is paid to the effect of Pu as fuel inside these CPs thereby aiming to indicate which mechanisms are of concern when such CPs are considered as fuel in future reactors. In the last part of this chapter constraints are listed that were imposed to the models considered in the framework of this thesis. Chapter 3 presents the results of unit-cell calculations performed with three code systems. The main objective of this chapter is to compare the calculational results of one particular code system, which is a candidate for the generation of cross sections for a full-core calculation, to those of the other two code systems. Also some reactor physics interpretations of the calculational results are presented. The unit-cell calculations embrace the computation of a number of reactor physics parameters for pebbles with a varying plutonium mass per pebble and with different types of coated particles. For one pebble configuration, these parameters have been calculated for various fuel temperatures and over-all (uniform) temperatures. For that particular pebble configuration, also the results of a two burnup calculations were compared. Chapter 4 reports the results of a parameter study in which the number of coated particles per pebble as well as the type and size of the CPs have been varied. The effect of different pebble configurations on several reactor physics

  13. ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

    International Nuclear Information System (INIS)

    Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

    1977-01-01

    ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

  14. High temperature fusion reactor design

    International Nuclear Information System (INIS)

    Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

    1979-01-01

    Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

  15. System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

    International Nuclear Information System (INIS)

    Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

    2010-01-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  16. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2016-06-15

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  17. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    International Nuclear Information System (INIS)

    Konarek, E.; Coulas, B.; Sarvinis, J.

    2016-01-01

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  18. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  19. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  20. A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

    International Nuclear Information System (INIS)

    Luo, Chending; Zhao, Fuqiang; Zhang, Na

    2014-01-01

    Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

  1. The high-temperature reactor

    International Nuclear Information System (INIS)

    Kirchner, U.

    1991-01-01

    The book deals with the development of the German high-temperature reactor (pebble-bed), the design of a prototype plant and its (at least provisional) shut-down in 1989. While there is a lot of material on the HTR's competitor, the fast breeder, literature is very incomplete on HTRs. The author describes HTR's history as a development which was characterised by structural divergencies but not effectively steered and monitored. There was no project-oriented 'community' such as there was for the fast breeder. Also, the new technology was difficult to control there were situations where no one quite knew what was going on. The technical conditions however were not taken as facts but as a basis for interpretation, wishes and reservations. The HTR gives an opportunity to consider the conditions under which large technical projects can be carried out today. (orig.) [de

  2. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

  3. Mathematical modeling of a fluidized bed gasifier for steam gasification of coal using high-temperature nuclear reactor heat

    International Nuclear Information System (INIS)

    Kubiak, H.; vanHeek, K.-H.; Juntgen, H.

    1986-01-01

    Coal gasification is a well-known technique and has already been developed and used since a long time. In the last few years, forced by the energy situation, new efforts have been made to improve known processes and to start new developments. Conventional gasification processes use coal not only as feedstock to be gasified but also for supply of energy for reaction heat, steam production, and other purposes. With a nuclear high temperature reactor (HTR) as a source for process heat, it is possible to transform the whole of the feed coal into gas. This concept offers advantages over existing gasification processes: saving of coal, as more gas can be produced from coal; less emission of pollutants, as the HTR is used for the production of steam and electricity instead of a coal-fired boiler; and lower production costs for the gas

  4. The study on the role of very high temperature reactor and nuclear process heat utilization in future energy systems

    International Nuclear Information System (INIS)

    Yasukawa, Sigeru; Mankin, Shuichi; Tadokoro, Yoshihiro; Sato, Osamu; Yamaguchi, Kazuo; Ueno, Seiichi

    1986-11-01

    This report describes the analytical results being made in the study on the role of Very High Temperature Reactor and nuclear process heat utilization in future energy system, which is aimed at zero emission. In the former part of the report, the modeling of the reference energy system, main characteristics of energy technologies, and scenario indicators as well as system behavioral objectives for optimization are explained. In the latter part, analytical results such as the time-period variation of overall energy utilization efficiency, energy supply/demand structure in long-terms, energy contribution and economic competition of new energy technologies, environmental effluents released through verious energy activities, impacts to and from national economy, and some sensitivity analyses, are reviewed. (author)

  5. A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Du Toit, C.G.; Rousseau, P.G.; Greyvenstein, G.P.; Landman, W.A.

    2006-01-01

    The theoretical basis and conceptual formulation of a comprehensive reactor model to simulate the thermal-fluid phenomena of the PBMR reactor core and core structures is given. Through a rigorous analysis the fundamental equations are recast in a form that is suitable for incorporation in a systems CFD code. The formulation of the equations results in a collection of one-dimensional elements (models) that can be used to construct a comprehensive multi-dimensional network model of the reactor. The elements account for the pressure drop through the reactor; the convective heat transport by the gas; the convection heat transfer between the gas and the solids; the radiative, contact and convection heat transfer between the pebbles and the heat conduction in the pebbles. Results from the numerical model are compared with that of experiments conducted on the SANA facility covering a range of temperatures as well as two different fluids and different heating configurations. The good comparison obtained between the simulated and measured results show that the systems CFD approach sufficiently accounts for all of the important phenomena encountered in the quasi-steady natural convection driven flows that will prevail after critical events in a reactor. The fact that the computer simulation time for all of the simulations was less than three seconds on a standard notebook computer also indicates that the new model indeed achieves a fine balance between accuracy and simplicity. The new model can therefore be used with confidence and still allow quick integrated plant simulations. (authors)

  6. Stability analysis of the high temperature thermal pebble bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1981-02-01

    A study was made of the stability of the high temperature gas-cooled pebble bed core against xenon-driven oscillation. This generic study indicated that a core as large as 3000 MW(t) could be stable. Several aspects present a challenge to analysis including the void space above the pebble bed, the effects of possible control rod configurations, and the temperature feedback contribution. Special methods of analysis were developed in this effort. Of considerable utility was the scheme of including an azimuthal buckling loss term in the neturon balance equations admitting direct solution of the first azimuthal harmonic for a core having azimuthal symmetry. This technique allows the linear stability analysis to be done solving two-dimensional (RZ) problems instead of three-dimensional problems. A scheme for removing the fundamental source contribution was also implemented to allow direct iteration toward the dominant harmonic solution, treating up to three dimensions with diffusion theory

  7. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Battaglia, Francine

    2008-01-01

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  8. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  9. High-temperature reactor in modular construction

    International Nuclear Information System (INIS)

    Mueller, F.U.; Reutler, H.; Ullrich, M.

    1981-01-01

    Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

  10. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  11. Experimental investigation of high temperature high voltage thermionic diode for the space power nuclear reactor

    International Nuclear Information System (INIS)

    Onufriyev, Valery V.

    2001-01-01

    It is well known that the rise of arc from the dense glow discharge is connected with the thermion and secondary processes on the cathode surface (Granovsky, 1971; Leob, 1953; Engel, 1935). First model of breakdown of the cathode layer is connected with the increase of the cathode temperature in consequence of the ion bombardment that leads to the grows its thermo-emissive current. Other model shows the main role of the secondary effects on the cathode surface-the increase of the secondary ion emission coefficient--γ i with the grows of glow discharge voltage. But the author of this investigation work of breakdown in Cs vapor (a transmission the glow discharge into self-maintaining arc discharge) discovered the next peculiarity: the value of breakdown voltage is constant when the values of vapor temperature (its pressure p cs ) and cathode temperature T k is constant too (U b =constant with T k =constant and p cs =constant) and it is not a statistical value (Onufryev, Grishin, 1996) (that was observed in gas glow discharges other authors (Granovsky, 1971; Leob, 1953; Engel, 1935)). The investigations of thermion high voltage high temperature diode (its breakdown characteristics in closed state and voltage-current characteristics in disclosed state) showed that the value of the breakdown voltage is depended on the vapor pressure in inter-electrode gap (IEG)-p cs and cathode temperature-T k and is independent on IEG length--Δ ieg . On this base it was settled that the main role in transition of glow discharge to self-maintaining arc discharge plays an ion cathode layer but more exactly--the region of excited atoms--''Aston glow.''

  12. Experimental investigation of high temperature high voltage thermionic diode for the space power nuclear reactor

    Science.gov (United States)

    Onufriyev, Valery. V.

    2001-02-01

    It is well known that the rise of arc from the dense glow discharge is connected with the thermion and secondary processes on the cathode surface (Granovsky, 1971; Leob, 1953; Engel, 1935). First model of breakdown of the cathode layer is connected with the increase of the cathode temperature in consequence of the ion bombardment that leads to the grows its thermo-emissive current. Other model shows the main role of the secondary effects on the cathode surface-the increase of the secondary ion emission coefficient-γi with the grows of glow discharge voltage. But the author of this investigation work of breakdown in Cs vapor (a transmission the glow discharge into self-maintaining arc discharge) discovered the next peculiarity: the value of breakdown voltage is constant when the values of vapor temperature (its pressure pcs) and cathode temperature Tk is constant too (Ub=constant with Tk=constant and pcs=constant) and it is not a statistical value (Onufryev, Grishin, 1996) (that was observed in gas glow discharges other authors (Granovsky, 1971; Leob, 1953; Engel, 1935)). The investigations of thermion high voltage high temperature diode (its breakdown characteristics in closed state and voltage-current characteristics in disclosed state) showed that the value of the breakdown voltage is depended on the vapor pressure in inter-electrode gap (IEG)-pcs and cathode temperature-Tk and is independent on IEG length-Δieg. On this base it was settled that the main role in transition of glow discharge to self-maintaining arc discharge plays an ion cathode layer but more exactly-the region of excited atoms-``Aston glow.'' .

  13. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  14. Nuclear fuels for very high temperature applications

    International Nuclear Information System (INIS)

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO 2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures

  15. Thermodynamic study of residual heat from a high temperature nuclear reactor to analyze its viability in cogeneration processes

    International Nuclear Information System (INIS)

    Santillan R, A.; Valle H, J.; Escalante, J. A.

    2015-09-01

    In this paper the thermodynamic study of a nuclear power plant of high temperature at gas turbine (GTHTR300) is presented for estimating the exploitable waste heat in a process of desalination of seawater. One of the most studied and viable sustainable energy for the production of electricity, without the emission of greenhouse gases, is the nuclear energy. The fourth generation nuclear power plants have greater advantages than those currently installed plants; these advantages have to do with security, increased efficiencies and feasibility to be coupled to electrical cogeneration processes. In this paper the thermodynamic study of a nuclear power plant type GTHTR300 is realized, which is selected by greater efficiencies and have optimal conditions for use in electrical cogeneration processes due to high operating temperatures, which are between 700 and 950 degrees Celsius. The aim of the study is to determine the heat losses and the work done at each stage of the system, determining where they are the greatest losses and analyzing in that processes can be taken advantage. Based on the study was appointed that most of the energy losses are in form of heat in the coolers and usually this is emitted into the atmosphere without being used. From the results a process of desalination of seawater as electrical cogeneration process is proposed. This paper contains a brief description of the operation of the nuclear power plant, focusing on operation conditions and thermodynamic characteristics for the implementation of electrical cogeneration process, a thermodynamic analysis based on mass and energy balance was developed. The results allow quantifying the losses of thermal energy and determining the optimal section for coupling of the reactor with the desalination process, seeking to have a great overall efficiency. (Author)

  16. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  17. Nuclear reactor pressure vessel with an inner metal coating covered with a high temperature resistant thermal insulator

    International Nuclear Information System (INIS)

    1974-01-01

    The thermal insulator covering the metal coating of a reactor vessel is designed for resisting high temperatures. It comprises one or several porous layers of ceramic fibers or of stacked metal foils, covered with a layer of bricks or ceramic tiles. The latter are fixed in position by fasteners comprising pins fixed to the coating and passing through said porous layers and fasteners (nut or bolts) for individually fixing the bricks to said pins, whereas ceramic plugs mounted on said bricks or tiles provide for the thermal insulation of the pins and of the nuts or bolts; such a thermal insulation can be applied to high-temperature reactors or to fast reactors [fr

  18. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    Science.gov (United States)

    Shchelik, S. V.; Pavlov, A. S.

    2013-07-01

    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  19. Resonance integral calculations for high temperature reactors

    International Nuclear Information System (INIS)

    Blake, J.P.H.

    1960-02-01

    Methods of calculation of resonance integrals of finite dilution and temperature are given for both, homogeneous and heterogeneous geometries, together with results obtained from these methods as applied to the design of high temperature reactors. (author)

  20. Development of Very High Temperature Reactor Technology

    International Nuclear Information System (INIS)

    Lee, Won Jae; Noh, J. M.; Kim, Y. H.

    2009-04-01

    For an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

  1. High Temperature reactors status 1977

    International Nuclear Information System (INIS)

    Ekholm, R.; Bosaeus, J.; Carleson, G.; Gelin, R.; Jirlow, K.; Linder, S.; Menon, S.; Runfors, U.; Vieider, G.

    1978-03-01

    The objective of this report is to summarize the current state-of-the-art of HTR technology as part of follow-up studies of the development of advanced fission reactor systems. These studies have been performed at AB Atomenergi since fiscal year 1975/76 and are financed by governmental funds for energy R and D. In this report emphasis is given to the following main aspects of the HTR development: - a survey of the major HTR - R and D programmes; - the description of HTR technology including remaining development problems and uncertainties; - the analysis of the safety and environmental characteristics of the HTR systems; - the analysis of the incentives for the introduction of various HTR types. The report contains also information kindly provided directly by experts from several organisations developing the HTR-systems

  2. The study on the role of very high temperature reactor and nuclear process heat utilization in future energy systems

    International Nuclear Information System (INIS)

    Yasukawa, Shigeru; Mankin, Shuichi; Sato, Osamu; Tadokoro, Yoshihiro; Nakano, Yasuyuki; Nagano, Takao; Yamaguchi, Kazuo; Ueno, Seiichi.

    1987-11-01

    The objectives of the systems analysis study on ''The Role of High Temperature Nuclear Heat in Future Energy Systems'' under the cooperative research program between Japan Atomic Energy Research Institute and the Massachusetts Institute of Technology are to analyze the effect and the impact of introduction of high temperature nuclear heat in Japanese long-term energy systems aiming at zero environmental emissions from view points of energy supply/demand, economy progress, and environmental protection, and to show the potentials of involved technologies and to extract the associated problems necessary for research and developments. This report describes the results being obtained in these three years from 1985. The present status of our energy system are explained at first, then, our findings concerning on analytical approach, method for analysis, view points to the future, scenario state space, reference energy systems, evolving technologies in it, and results analyzed are described. (author)

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

    International Nuclear Information System (INIS)

    Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

    1974-08-01

    The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

  5. High temperature heat exchange: nuclear process heat applications

    International Nuclear Information System (INIS)

    Vrable, D.L.

    1980-09-01

    The unique element of the HTGR system is the high-temperature operation and the need for heat exchanger equipment to transfer nuclear heat from the reactor to the process application. This paper discusses the potential applications of the HTGR in both synthetic fuel production and nuclear steel making and presents the design considerations for the high-temperature heat exchanger equipment

  6. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  7. High temperature reactor safety and environment

    International Nuclear Information System (INIS)

    Brisbois, J.; Charles, J.

    1975-01-01

    High-temperature reactors are endowed with favorable safety and environmental factors resulting from inherent design, main-component safety margins, and conventional safety systems. The combination of such characteristics, along with high yields, prove in addition, that such reactors are plagued with few problems, can be installed near users, and broaden the recourse to specific power, therefore fitting well within a natural environment [fr

  8. High temperature brazing of reactor materials

    International Nuclear Information System (INIS)

    Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

    1990-01-01

    Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

  9. Production analysis of methanol and hydrogen of a modificated blast furnace gas using nuclear energy of the high temperature reactor

    International Nuclear Information System (INIS)

    Peschel, W.

    1985-12-01

    Modern blast furnaces are operated with a coke ration of 500 kg/t pig iron. The increase of the coke ratio to 1000 kg/t pig iron raises the content of carbon monoxide and hydrogen in the blast furnace gas. On the basis of a blast furnace gas modificated in such a way, the production of methanol and hydrogen is investigated under the coupling of current and process heat from the high temperature reactor. Moreover the different variants are discussed, for which respectively a material and energetic balance as well as an estimation of the production costs is performed. Regarding the subsequent treatment of the blast furnace gas it turns out favourably in principle to operate the blast furnace with a nitrogen-free wind consisting only of oxygen and steam. The production costs show a strong dependence on the raw material costs, whose influence is shown in a nomograph. (orig.) [de

  10. High temperature fast reactor for hydrogen production in Brazil; Reator nuclear rapido de altissima temperatura para producao de hidrogenio no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: jamil@ieav.cta.br

    2008-07-01

    The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, {approx} 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

  11. The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 10

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Woike, O.

    1974-08-01

    The necessary development steps, which have to be taken for the construction of a prototype plant for nuclear process heat, are enumerated. In particular, the work which is involved for the development of the nuclear steam-reforming technique, for the further development of the ball-shaped fuel elements at high gas outlet temperatures and for the reactor components, is described in detail. A brief survey of the needs of development of the IHX (intermediate heat exchanger) is given. An attempt is made to give overall time and cost estimates. (orig.) [de

  12. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  13. Evaluation of two processes of hydrogen production starting from energy generated by high temperature nuclear reactors; Evaluacion de dos procesos de produccion de hidrogeno a partir de energia generada por reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J., E-mail: jvalle@upmh.edu.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard Acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2013-10-15

    In this work an evaluation to two processes of hydrogen production using energy generated starting from high temperature nuclear reactors (HTGR's) was realized. The evaluated processes are the electrolysis of high temperature and the thermo-chemistry cycle Iodine-Sulfur. The electrolysis of high temperature, contrary to the conventional electrolysis, allows reaching efficiencies of up to 60% because when increasing the temperature of the water, giving thermal energy, diminishes the electric power demand required to separate the molecule of the water. However, to obtain these efficiencies is necessary to have water vapor overheated to more than 850 grades C, temperatures that can be reached by the HTGR. On the other hand the thermo-chemistry cycle Iodine-Sulfur, developed by General Atomics in the 1970 decade, requires two thermal levels basically, the great of them to 850 grades C for decomposition of H{sub 2}SO{sub 4} and another minor to 360 grades C approximately for decomposition of H I, a high temperature nuclear reactor can give the thermal energy required for the process whose products would be only hydrogen and oxygen. In this work these two processes are described, complete models are developed and analyzed thermodynamically that allow to couple each hydrogen generation process to a reactor HTGR that will be implemented later on for their dynamic simulation. The obtained results are presented in form of comparative data table of each process, and with them the obtained net efficiencies. (author)

  14. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

  15. Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

    2008-01-01

    The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

  16. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  17. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  18. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    International Nuclear Information System (INIS)

    Phillpot, Simon; Tulenko, James

    2011-01-01

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  19. Very high temperature measurements: Applications to nuclear reactor safety tests; Mesures des tres hautes temperatures: Applications a des essais de surete des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Parga, Clemente-Jose

    2013-09-27

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  20. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    International Nuclear Information System (INIS)

    Pauzi, A M

    2013-01-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  1. Modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1988-01-01

    The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

  2. Materials for high temperature reactor vessels

    International Nuclear Information System (INIS)

    Buenaventura Pouyfaucon, A.

    2004-01-01

    Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

  3. A high temperature reactor for ship propulsion

    International Nuclear Information System (INIS)

    Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

    2002-01-01

    The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

  4. High temperature nuclear heat for isothermal reformer

    International Nuclear Information System (INIS)

    Epstein, M.

    2000-01-01

    High temperature nuclear heat can be used to operate a reformer with various feedstock materials. The product synthesis gas can be used not only as a source for hydrogen and as a feedstock for many essential chemical industries, such as ammonia and other products, but also for methanol and synthetic fuels. It can also be burnt directly in a combustion chamber of a gas turbine in an efficient combined cycle and generate electricity. In addition, it can be used as fuel for fuel cells. The reforming reaction is endothermic and the contribution of the nuclear energy to the calorific value of the final product (synthesis gas) is about 25%, compared to the calorific value of the feedstock reactants. If the feedstock is from fossil origin, the nuclear energy contributes to a substantial reduction in CO 2 emission to the atmosphere. The catalytic steam reforming of natural gas is the most common process. However, other feedstock materials, such as biogas, landfill gas and CO 2 -contaminated natural gas, can be reformed as well, either directly or with the addition of steam. The industrial steam reformers are generally fixed bed reactors, and their performance is strongly affected by the heat transfer from the furnace to the catalyst tubes. In top-fired as well as side-fired industrial configurations of steam reformers, the radiation is the main mechanism of heat transfer and convection heat transfer is negligible. The flames and the furnace gas constitute the main sources of the heat. In the nuclear reformers developed primarily in Germany, in connection with the EVA-ADAM project (closed cycle), the nuclear heat is transferred from the nuclear reactor coolant gas by convection, using a heating jacket around the reformer tubes. In this presentation it is proposed that the helium in a secondary loop, used to cool the nuclear reactor, will be employed to evaporate intermediate medium, such as sodium, zinc and aluminum chloride. Then, the vapors of the medium material transfer

  5. Preliminary risk analysis of an Hydrogen production plant using the reformed process of methane with vapor coupled to a high temperature nuclear reactor

    International Nuclear Information System (INIS)

    Flores y Flores, A.; Nelson E, P.F.; Francois L, J.L.

    2004-01-01

    It is necessary to identify the different types of dangers, as well as their causes, probabilities and consequences of the same ones, inside plants, industries and any process to classify the risks. This work is focused in particular to a study using the technical HAZOP (Hazard and Operability) for a plant of reformed of methane with vapor coupled to a nuclear reactor of the type HTTR (High Temperature Test Reactor), which is designed to be built in Japan. In particular in this study the interaction is analyzed between the nuclear reactor and the plant of reformed of methane with vapor. After knowing the possible causes of risk one it is built chart of results of HAZOP to have a better vision of the consequences of this faults toward the buildings and constructions, to people and the influence of the fault on each plant; for what there are proposed solutions to mitigate these consequences or to avoid them. The work is divided in three sections: a brief introduction about the technique of HAZOP; some important aspects of the plant of reformed of methane with vapor; and the construction of the chart of results of HAZOP. (Author)

  6. Development history of the gas turbine modular high temperature reactor

    International Nuclear Information System (INIS)

    Brey, H.L.

    2001-01-01

    The development of the high temperature gas cooled reactor (HTGR) as an environmentally agreeable and efficient power source to support the generation of electricity and achieve a broad range of high temperature industrial applications has been an evolutionary process spanning over four decades. This process has included ongoing major development in both the HTGR as a nuclear energy source and associated power conversion systems from the steam cycle to the gas turbine. This paper follows the development process progressively through individual plant designs from early research of the 1950s to the present focus on the gas turbine modular HTGR. (author)

  7. Power Conversion Study for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Richard Moore; Robert Barner

    2005-01-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

  8. Materials for advanced high temperature reactors

    International Nuclear Information System (INIS)

    Graham, L.W.

    1977-01-01

    Materials are studied in advanced applications of high temperature reactors: helium gas turbine and process heat. Long term creep behavior and corrosion tests are conducted in simulated HTR helium up to 1000 deg C with impurities additions in the furnace atmosphere. Corrosion studies on AISI 321 steels at 800-1000 deg C have shown that the O 2 partial pressure is as low as 10 -24+-3 atm, Ni and Fe cannot be oxidised above about 500 and 600 deg C, Cr cease to oxidise at 800 to 900 deg C and Ti at 900 to 1000 deg C depending on alloy composition γ' strengthened superalloys must depend on a protective corrosion mechanism assisted by the presence of Ti and possibly Cr. Carburisation has been identified metallographically in several high temperature materials: Hastelloy X and M21Z. Alloy TZM appears to be inert in HTR Helium at 900 and 1000 deg C. In alloy 800 and Inconel 625 surface cracks initiation is suppressed but crack propagation is accelerated but this was not apparent in AISI steels, Hastelloy X or fine grain Inconel at 750 deg C

  9. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  10. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

  11. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    International Nuclear Information System (INIS)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

    2015-01-01

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO 2 emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO 2 emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus TM Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition

  12. High-temperature Gas Reactor (HTGR)

    Science.gov (United States)

    Abedi, Sajad

    2011-05-01

    General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

  13. Evaluation of the security of a hydrogen producing plant by means of the S I cycle coupled to a nuclear reactor of high temperature

    International Nuclear Information System (INIS)

    Ruiz S, T.; Francois, J. L.; Nelson, P. F.; Cruz G, M. J.

    2011-11-01

    At the present one of the processes that demonstrates, theoretically, to be one of the most efficient for the hydrogen production is the thermal-chemistry cycle Sulfur-Iodine. One way of obtaining the temperature ranges required by the process is through the helium coming from a very high temperature reactor. The coupling of the chemical plant with the nuclear plant presents aspects of security that should be analyzed; among them the analysis of the danger of the process materials is, with the purpose of implementing security measures to protect the facilities and equipment s, the environment and the population. These measures can be: emergency answer plans of the stations, definition of the minimum distance required among facilities, determination of the exclusion area, etc. In this study simulations were made with the computer code Phast in order to knowing the possible affectation areas due to the liberation of a great quantity of energy due to a helium leak to very high temperature, of toxic materials or by a possible hydrogen combustion. The results for the liberations of sulfuric acid, hydrogen, iodine, helium and sulfur dioxide are shown, specially. The operation conditions were taken of a combination of the preliminary design proposed by General Atomics and the optimized conditions by the Korea Advanced Institute of Science and Technology, considering a production of 1 kg-mol/s of hydrogen. The iodine was the material that presented a major affectation area. (Author)

  14. Proposal for the small high temperature gas cooled reactor (application of nuclear energy in the changing and emerging energy markets)

    International Nuclear Information System (INIS)

    Remond R Pahladsingh

    2005-01-01

    With the peaking of the oil-production (2006) on the global energy-markets, already predicted by M.K. Hubbert in 1970 and the limited possibilities to use coal as an alternative, the world has come in a critical situation. Exponential growth in global population and the demand of energy (IEA) have created a unique, but known, problem in the present time. The difference with the past is the period in which the changes will take place is not centuries or decades but most likely a few years. There is no alternative for oil to anticipate on the coming energy-crisis. The coal-industry has been demolished in Europe and America, under the pressure of the environmental lobby, and the nuclear industry is in very bad shape. Large scale nuclear plants can only be built in strong power grids; the present power grids are exposed to de-investment and less maintenance in liberalized and de-regulated economies. Furthermore the Digital Society is placing additional requirements on reliability and quality of our electric power systems. Life extension programs in the nuclear industry have shown that the present industry is robust, safety standards are high but at the same time no breakthrough technology has come to the market. The manpower in the nuclear industry is ageing. There are no real indications of growth in the large scale nuclear industry. The present industry is heavily relaying on the present grid concepts that have shown many difficulties in recent past and the increased failures in power-grids will create constraints to excel the option of nuclear energy in the electric power industry This paper is a request to the nuclear industry to come with a revival program, easy to implement nuclear technology and to help the world to solve the huge energy-problems that present and future generations are facing in the world. The proposal in the paper is a request to bring nuclear energy back in our society (Yes in my backyard-YIMBY) so that the many fruitful uses of nuclear energy

  15. Summary - Advanced high-temperature reactor for hydrogen and electricity production

    International Nuclear Information System (INIS)

    Forsberg, Charles W.

    2001-01-01

    Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

  16. Very-high-temperature reactors for future use

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1988-08-01

    Very-high-temperature reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at reactor coolant temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

  17. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Tsuchie, Yasuo; Kunitomi, Kazuhiko; Shiozawa, Shusaku; Konuki, Kaoru; Inagaki, Yoshiyuki; Hayakawa, Hitoshi

    2002-01-01

    The high temperature gas-cooled reactor (HTGR) is now under a condition difficult to know it well, because of considering not only power generation, but also diverse applications of its nuclear heat, of having extremely different safe principle from that of conventional reactors, of having two types of pebble-bed and block which are extremely different types, of promoting its construction plan in South Africa, of including its application to disposition of Russian surplus weapons plutonium of less reporting HTTR in Japan in spite of its full operation, and so on. However, HTGR is expected for an extremely important nuclear reactor aiming at the next coming one of LWR. HTGR which is late started and developed under complete private leading, is strongly conscious at environmental problem since its beginning. Before 30 years when large scale HTGR was expected to operate, it advertised a merit to reduce wasted heat because of its high temperature. As ratio occupied by electricity expands among application of energies, ratio occupied by the other energies are larger. When considering applications except electric power, high temperature thermal energy from HTGR can be thought wider applications than that from LWR and so on. (G.K.)

  18. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  19. Mechanical properties of concrete for power reactor at high temperatures

    International Nuclear Information System (INIS)

    Kawase, Kiyotaka; Tanaka, Hitoshi; Nakano, Masayuki

    1985-01-01

    The purpose of this study is to investigate the mechanical properties of concrete for power reactor at high temperature. This paper presents the creep behavior of concrete at high temperature and the cause by which a specified aggregate is broken at a specified high temperature. The creep coefficient at high temperature is smaller than that at ordinary temperature. (author)

  20. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  1. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D; Estudo termofluidodinâmico de reatores nucleares avançados de alta temperatura utilizando o RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Scari, Maria Elizabeth

    2017-07-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF{sub 2}, the LiF-BeF{sub 2}, also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO

  2. Gas cooled thermal reactors with high temperatures (VHTR)

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.

    2014-01-01

    VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)

  3. High temperature fast reactor for hydrogen production in Brazil

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

    2008-01-01

    The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, ∼ 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

  4. The high-temperature reactor's attractiveness lies in passive safety

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    In the recent years the use of nuclear energy has turned from a technical and scientific issue to a political one. The high-temperature reactor (HTR) however, has always been advertised as particularly safe. The present situation and future developments of HTR-technology were the two issues that VDI-News brought up on the 27th October on an HTR-conference in an interview with the 'spiritual father' of the HTR, Prof. Dr. Rudolf Schulten of the Juelich Nuclear Research Centre. (orig.) [de

  5. Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

    International Nuclear Information System (INIS)

    Elsheikh, B.M.

    2017-01-01

    Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

  6. Fusion reactors-high temperature electrolysis (HTE)

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800 0 C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400 0 C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H 2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10 6 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

  7. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  8. Nuclear and quark matter at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Biro, Tamas S. [H.A.S. Wigner Research Centre for Physics, Budapest (Hungary); Jakovac, Antal [Roland Eotvos University, Budapest (Hungary); Schram, Zsolt [University of Debrecen, Institute for Theoretical Physics, Debrecen (Hungary)

    2017-03-15

    We review important ideas on nuclear and quark matter description on the basis of high-temperature field theory concepts, like resummation, dimensional reduction, interaction scale separation and spectral function modification in media. Statistical and thermodynamical concepts are spotted in the light of these methods concentrating on the -partially still open- problems of the hadronization process. (orig.)

  9. Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

    International Nuclear Information System (INIS)

    Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

    2012-01-01

    Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  11. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  12. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

    Chikamatsu N, M. A.; Puente E, F.

    2014-10-01

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  13. Use of thorium for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

  14. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented concerning the Dragon reactor support studies and fuel irradiation programs, HTGR and fuel graphite studies, primary circuit materials, reactor safety evaluation, and administration

  15. Process flow sheet evaluation of a nuclear hydrogen steelmaking plant applying very high temperature reactors for efficient steel production with less CO{sub 2} emissions

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Seiji, E-mail: kasahara.seiji@jaea.go.jp; Inagaki, Yoshiyuki; Ogawa, Masuro

    2014-05-01

    Highlights: • CO{sub 2} emissions from a nuclear hydrogen steelmaking system was 13–21% of that from a blast furnace steelmaking system. • Heat input to shaft furnace in hydrogen steelmaking was large with much H{sub 2} consumption in the part. • Though hydrogen production thermal efficiency had influence on total heat input to hydrogen steelmaking, the effect on the CO{sub 2} emissions was small. • Steelmaking scale of a nuclear hydrogen steelamking plant with 2 VHTRs was a little smaller than that of the largest Midrex{sup ®} steelmaking plants. - Abstract: Recently, CO{sub 2} reduction is an important problem for steelmaking. Substitution of coal, presently used as a reducing agent of iron ore in blast furnaces, to hydrogen produced by non-fossil energy is a way to reduce CO{sub 2} emissions. In this study, the idea of nuclear hydrogen steelmaking (NHS) system was investigated using very high temperature reactor (VHTR) and thermochemical hydrogen production iodine–sulfur (IS) process. Heat input and CO{sub 2} emissions including material production, material transportation, and electricity generation were evaluation criteria. Results of the NHS system were compared with those of a conventional blast furnace steelmaking (BFS) system. Influence of heat input options to the steelmaking process and hydrogen production thermal efficiency of IS process were investigated for the NHS system. Though heat input to the NHS system was 130–142% of that to the BFS system, CO{sub 2} emissions of the system were 13–21%. Pre-heating of hydrogen by coal combustion before blowing to a shaft furnace was effective to decrease heat input, although CO{sub 2} emissions increased. Direct pre-heating by nuclear heat was also effective without increase of CO{sub 2} emissions if close location of the nuclear reactor to the steelmaking plant was publicly accepted. Hydrogen production thermal efficiency had a significant influence on the heat input. Conceptual design of a

  16. Assessment of very high temperature reactors in process applications

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

    1976-01-01

    In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

  17. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  18. Very-high-temperature reactors for future use

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1988-01-01

    Very-High-Temperature Reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at helium temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  20. High temperature resistant materials and structural ceramics for use in high temperature gas cooled reactors and fusion plants

    International Nuclear Information System (INIS)

    Nickel, H.

    1992-01-01

    Irrespective of the systems and the status of the nuclear reactor development lines, the availability, qualification and development of materials are crucial. This paper concentrates on the requirements and the status of development of high temperature metallic and ceramic materials for core and heat transferring components in advanced HTR supplying process heat and for plasma exposed, high heat flux components in Tokamak fusion reactor types. (J.P.N.)

  1. Hydrogen production from high temperature electrolysis and fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

    1978-01-01

    Production of hydrogen from high temperature electrolysis of steam coupled with a fusion reactor is studied. The process includes three major components: the fusion reactor, the high temperature electrolyzer and the power conversion cycle each of which is discussed in the paper. Detailed process design and analysis of the system is examined. A parametric study on the effect of process efficiency is presented

  2. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  3. Thermodynamic study of residual heat from a high temperature nuclear reactor to analyze its viability in cogeneration processes; Estudio termodinamico del calor residual de un reactor nuclear de alta temperatura para analizar su viabilidad en procesos de cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Santillan R, A.; Valle H, J.; Escalante, J. A., E-mail: santillanaura@gmail.com [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2015-09-15

    In this paper the thermodynamic study of a nuclear power plant of high temperature at gas turbine (GTHTR300) is presented for estimating the exploitable waste heat in a process of desalination of seawater. One of the most studied and viable sustainable energy for the production of electricity, without the emission of greenhouse gases, is the nuclear energy. The fourth generation nuclear power plants have greater advantages than those currently installed plants; these advantages have to do with security, increased efficiencies and feasibility to be coupled to electrical cogeneration processes. In this paper the thermodynamic study of a nuclear power plant type GTHTR300 is realized, which is selected by greater efficiencies and have optimal conditions for use in electrical cogeneration processes due to high operating temperatures, which are between 700 and 950 degrees Celsius. The aim of the study is to determine the heat losses and the work done at each stage of the system, determining where they are the greatest losses and analyzing in that processes can be taken advantage. Based on the study was appointed that most of the energy losses are in form of heat in the coolers and usually this is emitted into the atmosphere without being used. From the results a process of desalination of seawater as electrical cogeneration process is proposed. This paper contains a brief description of the operation of the nuclear power plant, focusing on operation conditions and thermodynamic characteristics for the implementation of electrical cogeneration process, a thermodynamic analysis based on mass and energy balance was developed. The results allow quantifying the losses of thermal energy and determining the optimal section for coupling of the reactor with the desalination process, seeking to have a great overall efficiency. (Author)

  4. High temperature gas-cooled reactor: gas turbine application study

    International Nuclear Information System (INIS)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project

  5. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  6. Studies on high temperature research reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yuanhui; Zuo Kanfen [Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)

    1999-08-01

    China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

  7. Studies on high temperature research reactor in China

    International Nuclear Information System (INIS)

    Xu Yuanhui; Zuo Kanfen

    1999-01-01

    China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

  8. Nuclear heat for high temperature fossil fuel processing

    International Nuclear Information System (INIS)

    Walton, G.N.

    1981-01-01

    This is a report of a one-day symposium held at the Royal Institution, London, on 28 April 1981. It was organized by the Institute of Energy (London and Home Counties section) under the chairmanship of Dr A M Brown with the assistance of the Institute of Energy's Nuclear Special Interest Group. The following five papers were presented (available as a booklet, from the Institute of Energy, price Pound12.00): 1) The Dragon project and the High Temperature Reactor (HTR) position. Dr L Shepherd, UKAEA, Winfrith. 2) Coal gasification technology. Dr M St J Arnold, NCB, Stoke Orchard Laboratories. 3) The utilization of nuclear energy for coal gasification. Dr K H van Heek, G Hewing, R Kirchhoff and H J Schroter, Bergbau Forschung, Essen, West Germany. 4) The hydrogen economy. K F Langley, Energy Technology Support Unit, Harwell. 5) Economic perspectives and high temperature reactors. J D Thorn, director, Technical Services and Planning, UKAEA. (author)

  9. High temperature nuclear process heat systems for chemical processes

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.

    1976-01-01

    The development planning and status of the very high temperature gas cooled reactor as a source of industrial process heat is presented. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system offers a unique combination of the two that is environmentally and economically attractive and technically sound. Conceptual studies of several energy-intensive processes coupled to a nuclear heat source are presented

  10. Physics of high-temperature reactors

    International Nuclear Information System (INIS)

    Massimo, L.

    1976-01-01

    The subject is covered in chapters entitled: general description of the HTR core; general considerations about reactor physics; neutron cross-sections; basic aspects of transport and diffusion theory; methods for the solution of the diffusion equation; slowing-down and thermalization in graphite; resonance absorption; spectrum calculations and cross-section averaging; burn-up; core design; fuel management and cost calculations; temperature coefficient; core dynamics and accident analysis; reactor control; peculiarities of HTR physics; analysis of calculational accuracy; sequence of reactor design calculations. (U.K.)

  11. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  12. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  13. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

    In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

  14. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  15. High temperature reactor development in the Netherlands

    International Nuclear Information System (INIS)

    Heek, A.I. van

    1996-01-01

    This year, some clear design choices have been made in the WHITE Reactor development programme. The activities will be concentrated at the development of a small size pebble bed HTR for combined heat and power production with a closed cycle gas turbine. Objective of the development is threefold: 1. restoring social support; 2. establishing commercial viability after market introduction; and 3. making the market introduction itself feasible, i.e. limited development and first-of-a-kind costs. This design is based on the peu-a-peu design of KFA Juelich and will be optimized. The computer codes necessary for this are being prepared for this work. The dynamic neutronics code PANTHER is being coupled to the thermal hydraulics code THERMIX-DIREKT. For this reactor type, fuel temperatures are maximal in the scenario of depressurization with recriticality. Even for this scenario, fuel temperatures of the 20MWth PAP-GT do not exceed 1300 deg. C, so there should be room for upscaling for economic reasons. On the other hand, it would be convenient to fuel the reactor batchwise instead of continuously, and the use of thorium could be required. These two features may lead to a larger temperature margin. The optimal design must unite these features in the best acceptable way. To gain expertise in calculations on gas cooled graphite moderate reactors, benchmark calculations are being performed in parallel with international partners. Parallel to this, special expertise is being built up on HTR fuel and HTR reactor vessels. (author). 3 refs

  16. Process heat cogeneration using a high temperature reactor

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon; Valle, Edmundo del; Castillo, Rogelio

    2014-01-01

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU

  17. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Gustavo, E-mail: gustavoalonso3@gmail.com [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ramirez, Ramon [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Valle, Edmundo del [Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Castillo, Rogelio [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico)

    2014-12-15

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU.

  18. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  19. Programmes and projects for high-temperature reactor development

    International Nuclear Information System (INIS)

    Bogusch, Edgar; Hittner, Dominique

    2009-01-01

    An increasing attention has to be recognised worldwide on the development of High-Temperature Reactors (HTR) which has started in Germany and other countries in the 1970ies. While pebble bed reactors with spherical fuel elements have been developed and constructed in Germany, countries such as France, the US and Russia investigated HTR concepts with prismatic block-type fuel elements. The concept of a modular HTR formerly developed by Areva NP was an essential basis for the HTR-10 in China. A pebble bed HTR for electricity production is developed in South Africa. The construction is planned after the completion of the licensing procedure. Also the US is planning an HTR under the NGNP (Next Generation Nuclear Plant) Project. Due to the high temperature level of the helium coolant, the HTR can be used not only for electricity production but also for supply of process heat. Including its inherent safety features the HTR is an attractive candidate for heat supply to various types of plants e.g. for hydrogen production or coal liquefactions. The conceptual design of an HTR with prismatic fuel elements for the cogeneration of electricity and process heat has been developed by Areva NP. On the European scale the HTR development is promoted by the RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation) project. RAPHAEL is an Integrated Project of the Euratom 6th Framework Programme for the development of technologies towards a Very High-Temperature Reactor (VHTR) for the production of electricity and heat. It is financed jointly by the European Commission and the partners of the HTR Technology Network (HTR-TN) and coordinated by Areva NP. The RAPHAEL project not only promotes HTR development but also the cooperation with other European projects such as the material programme EXTREMAT. Furthermore HTR technology is investigated in the frame of Generation IV International Forum (GIF). The development of a VHTR with helium temperatures above 900 C for the

  20. The modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Lutz, D.E.; Lipps, A.J.

    1984-01-01

    Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

  1. Development of very high temperature reactor design technology

    International Nuclear Information System (INIS)

    Lee, Won Jae; Noh, Jan Man

    2012-04-01

    or an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, fission product/tritium transport analysis, core thermo-fluid analysis, system layout analysis, graphite structure seismic analysis and hydrogen exposion analysis, and they are being verified and validated through a lot of international collaborations

  2. High Temperature Electro-Mechanical Devices For Nuclear Applications

    International Nuclear Information System (INIS)

    Robertson, D.

    2010-01-01

    Nuclear power plants require a number of electro-mechanical devices, for example, Control Rod Drive Mechanisms (CRDM's) to control the raising and lowering of control rods and Reactor Coolant Pumps (RCP's) to circulate the primary coolant. There are potential benefits in locating electro-mechanical components in areas of the plant with high ambient temperatures. One such benefit is the reduced need to make penetrations in pressure vessels leading to simplified plant design and improved inherent safety. The feature that limits the ambient temperature at which most electrical machines may operate is the material used for the electrical insulation of the machine windings. Conventional electrical machines generally use polymer-based insulation that limits the ambient temperature they can operate in to below 200 degrees Celsius. This means that when a conventional electrical machine is required to operate in a hot area it must be actively cooled necessitating additional systems. This paper presents data gathered during investigations undertaken by Rolls-Royce into the design of high temperature electrical machines. The research was undertaken at Rolls-Royce's University Technology Centre in Advanced Electrical Machines and Drives at Sheffield University. Rolls- Royce has also been investigating high temperature wire and encapsulants and latterly techniques to provide high temperature insulation to terminations. Rolls-Royce used the experience gained from these tests to produce a high temperature electrical linear actuator at sizes representative of those used in reactor systems. This machine was tested successfully at temperatures equivalent to those found inside the reactor vessel of a pressurised water reactor through a full series of operations that replicated in service duty. The paper will conclude by discussing the impact of the findings and potential electro-mechanical designs that may utilise such high temperature technologies. (authors)

  3. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  4. High temperature gas cooled reactors in China

    International Nuclear Information System (INIS)

    He Jiachen; Qian Jihui

    1989-01-01

    China has plentiful energy resources, but it is unevenly distributed geographically. 60% of coal resources are concentrated in North China, 71% of hydro-power resources in the hardly accessible Southwest China, whereas the densely populated and highly industrialized 15 provinces/municipalities along the coast, yielding 73% of the gross national product, posses only 10% of national energy resources, which makes our railway system hard pressed. In fact, about 40% of the railway transport and 50% of the main waterway transport are committed to fuel. Yet the needs of energy in the coastal regions cannot be met. To develop nuclear power is a naturally expected approach to solving energy problems in China, particularly in the near term for the coastal regions, where the demand of electricity increases sharply and fuel transport from other regions is already tense. Chinese nuclear circle is interested in MHTGR due to the following reasons. 1. Small capacity of MHTGR is suitable for small power grid in certain areas. 2. Chinese manufacturers are able to provide whole package of conventional island of MHTGR nuclear power plant. 3. Multipurpose MHTGR is attractive for Chinese heavy industries. 4. MHTGR nuclear power plant can be built in suburbs due to inherent safety features. Regarding the users' requirements in China, it can be summarised as: 1. Mature technologies and easy to get license from nuclear safety authority. 2. Emergency zone as small as possible, even unnecessary. 3. 200-300 MWe size desirable. 4. Big portion of domestic share in engineering and component supply. 5. Slightly higher electricity price than coal fired. 6. Investment and favourable financing conditions from overseas. 7. Reimbursement of hard currency by countertrade. At present, four working groups, including users, manufacturers and nuclear industry circle, have been established for performing independent feasibility study on building MHTGR demonstration nuclear power plant in China. (author)

  5. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  6. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    2010-03-01

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  7. High-Temperature Reactor For Power Generation and District Heating

    International Nuclear Information System (INIS)

    Herzberger, Karlheinz

    1987-01-01

    The multinational BBC Brown Brave Group, which has its head-quarters in Baden/Switzerland, was founded in 1891. Its German company is Brown, Brave and CIEs AGM, Mannheim. The field of operation covers wide areas of electrical engineering: These includes mainly the manufacture of installations and equipment for the generation, conversion, distribution and utilization of electric power, with special emphasis on the capital goods sector. BBC erects turnkey power plants and manufactures electrical equipment for industrial plants and urban transport and main line trains. Also of major importance are standard electrical products such as motors, switches, cables, semiconductor devices as well as measuring and control equipment. In the field of nuclear power BBC is engaged in particular in the development and construction of high-temperature reactors for the generation of electric power and process heat. The following presentation gives a short view on the milestones of the HTR development achieved in 1987

  8. High-temperature-structural design and research and development for reactor system components

    International Nuclear Information System (INIS)

    Matsumura, Makoto; Hada, Mikio

    1985-01-01

    The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

  9. Design and development of gas turbine high temperature reactor 300

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Takizuka, Takakazu

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) has been designing a Japan's original gas turbine high temperature reactor, GTHTR300 (Gas Turbine High Temperature Reactor 300). The greatly simplified design based on salient features of the HTGR (High Temperature Gas-cooled reactor) with a closed helium gas turbine enables the GTHTR300 a high efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the HTTR (High Temperature Engineering Test Reactor) and fossil gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original features of this system are core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe, and the electric generation cost is close to a target cost of 4 Yen/kWh. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design except PCU design. Also, R and D for developing the power conversion unit is briefly described. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  10. High temperature reactors for cogeneration applications

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich (Germany). IEK-6; Allelein, Hans-Josef [Forschungszentrum Juelich (Germany). IEK-6; RWTH Aachen (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik (LRST)

    2016-05-15

    There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 C.

  11. HTR-2002: Proceedings of the conference on high temperature reactors

    International Nuclear Information System (INIS)

    2002-01-01

    High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

  12. High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

    International Nuclear Information System (INIS)

    Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

    2012-01-01

    Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

  13. The primary circuit of the dragon high temperature reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.

    2005-01-01

    The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

  14. Development of Very High Temperature Reactor Design Technology

    International Nuclear Information System (INIS)

    Lee, Won Jae; Noh, J. M.; Kim, K. S.

    2007-05-01

    To develop design technologies for the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature for an efficient hydrogen production, key studies were performed, which include evaluation technology for the performance and safety of VHTR, and development of design and analysis codes. First, to evaluate the performance of VHTR, a series of analyses has been carried out for core characteristics at 950 .deg. C, cooled-vessel adopting internal flow path through the graphite structure, compact heat exchanger with periodic channel configuration, intermediate loop system, risk/performance-informed method, and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC, safety evaluation of both VHTR and RCCS has been also performed. In addition, prototype codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

  15. High temperature electrolysis for hydrogen production using nuclear energy

    International Nuclear Information System (INIS)

    Herring, J. Stephen; O'brien, James E.; Stoots, Carl M.; Hawkes, Grant L.; Hartvigsen, Joseph J.

    2005-01-01

    High-temperature nuclear reactors have the potential for substantially increasing the efficiency of hydrogen production from water splitting, which can be accomplished via high-temperature electrolysis (HTE) or thermochemical processes. In order to achieve competitive efficiencies, both processes require high-temperature operation (∼850degC). High-temperature electrolytic water splitting supported by nuclear process heat and electricity has the potential to produce hydrogen with overall system efficiencies of 45 to 55%. At the Idaho National Laboratory, we are developing solid-oxide cells to operate in the steam electrolysis mode. The research program includes both experimental and modeling activities. Experimental results were obtained from ten-cell and 22-cell planar electrolysis stacks, fabricated by Ceramatec, Inc. The electrolysis cells are electrolyte-supported, with scandia-stabilized zirconia electrolytes (∼200 μm thick, 64 cm 2 active area), nickel-cermet steam/hydrogen electrodes, and manganite air-side electrodes. The metallic interconnect plates are fabricated from ferritic stainless steel. The experiments were performed over a range of steam inlet mole fractions, gas glow rates, and current densities. Hydrogen production rates greater than 100 normal liters per hour for 196 hours have been demonstrated. In order to evaluate the performance of large-scale HTE operations, we have developed single-cell models, based on FLUENT, and a process model, using the systems-analysis code HYSYS. (author)

  16. Optimization of the Neutronics of the Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    Zakova, Jitka; Talamo, Alberto

    2006-01-01

    In these studies, we have investigated the neutronic and safety performance of the Advanced High Temperature Reactor (AHTR) for plutonium and uranium fuels and we extended the analysis to five different coolants. The AHTR is a graphite-moderated and molten salt-cooled high temperature reactor, which takes advantage of the TRISO particles technology for the fuel utilization. The conceptual design of the core, proposed at the Oak Ridge National Laboratory, aims to provide an alternative to helium as coolant of high-temperature reactors for industrial applications like hydrogen production. We evaluated the influence of the radial reflector on the criticality of the core for the uranium and plutonium fuels and we focused on the void coefficient of 5 different molten salts; since the safety of the reactor is enhanced also by the large and negative coefficient of temperature, we completed our investigation by observing the keff changes when the graphite temperature varies from 300 to 1800 K. (authors)

  17. Secret high-temperature reactor concept for inertial fusion

    International Nuclear Information System (INIS)

    Monsler, M.J.; Meier, W.R.

    1983-01-01

    The goal of our SCEPTRE project was to create an advanced second-generation inertial fusion reactor that offers the potential for either of the following: (1) generating electricity at 50% efficiency, (2) providing high temperature heat (850 0 C) for hydrogen production, or (3) producing fissile fuel for light-water reactors. We have found that these applications are conceptually feasible with a reactor that is intrinsically free of the hazards of catastrophic fire or tritium release

  18. High temperature properties of nuclear reactor coolants and thermodynamic power cycle working fluids. Technical progress report, December 1, 1977--November 30, 1978

    International Nuclear Information System (INIS)

    Bonilla, C.F.; Holder, G.P.

    1978-09-01

    The program to determine the surface tension of sodium to high temperatures has been soundly commenced with the completion and calibration of a new all-molybdenum maximum-bubble-pressure apparatus suitable to some 1900 0 K. This property bears on boiling, condensing and two-phase flow phenomena, also related to LMFBR analysis. This same apparatus will be useful with lithium, of interest in fusion reactor cooling. Reduction of prior maximum bubble pressure results on potassium by the Schroedinger method has been repeated by the more precise but time-consuming Sugden procedure, which has increased the resulting surface tension values by an average of 0.27%. Thus the Sugden method will be employed to reduce the data for sodium and lithium

  19. Nuclear shell effects at high temperatures

    International Nuclear Information System (INIS)

    Davidson, N.J.; Miller, H.G.

    1993-01-01

    In discussing the disappearance of nuclear shell effects at high temperatures, it is important to distinguish between the ''smearing out'' of the single-particle spectrum with increasing temperature and the vanishing of shell related structures in many-body quantities such as the excitation energy per nucleon. We propose a semiempirical method to obtain an upper bound on the temperature required to smooth the single-particle spectrum, and point out that shell effects in many-body parameters may persist above this temperature. We find that the temperature required to smear out the single-particle spectrum is approximately 1 MeV for heavy nuclei (A approx-gt 150) and about 3--4 MeV for light nuclei (A approx-lt 50), in reasonable agreement with the estimate of 41/πA 1/3 obtained from calculations with harmonic oscillator potentials. These temperatures correspond to many-body excitation energies of approximately 20 and 60 MeV, respectively

  20. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  1. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  2. Modular high-temperature reactor launched (and wallchart)

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-01-01

    In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

  3. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  4. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  5. High-temperature turbopump assembly for space nuclear thermal propulsion

    Science.gov (United States)

    Overholt, David M.

    1993-01-01

    The development of a practical, high-performance nuclear rocket by the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program places high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio. The operating parameters arising from these goals drive the propellant-pump design. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is effected by rapid heating of the propellant from 100 K to thousands of degrees in the particle-bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. One approach to achieve high performance is to use an uncooled carbon-carbon nozzle and duct turbine inlet. The high-temperature capability is obtained by using carbon-carbon throughout the TPA hot section. Carbon-carbon components in development include structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines plus a wide variety of other turbomachinery applications.

  6. High-temperature turbopump assembly for space nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Overholt, D.M.

    1993-01-01

    The development of a practical, high-performance nuclear rocket by the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program places high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio. The operating parameters arising from these goals drive the propellant-pump design. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is effected by rapid heating of the propellant from 100 K to thousands of degrees in the particle-bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. One approach to achieve high performance is to use an uncooled carbon-carbon nozzle and duct turbine inlet. The high-temperature capability is obtained by using carbon-carbon throughout the TPA hot section. Carbon-carbon components in development include structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines plus a wide variety of other turbomachinery applications

  7. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW) [de

  8. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1978-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (orig./PW)

  9. Secondary heat exchanger design and comparison for advanced high temperature reactor

    International Nuclear Information System (INIS)

    Sabharwall, P.; Kim, E. S.; Siahpush, A.; McKellar, M.; Patterson, M.

    2012-01-01

    Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

  10. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  11. High-temperature reactors: a recent past, a near future

    International Nuclear Information System (INIS)

    Ballot, B.

    2007-01-01

    While high-temperature reactors did experience major developments in the past, in Europe in particular, significant R and D efforts are required, if a major innovation deployment is to be made possible, of modular reactors having the capability of being coupled, in reliable, economic fashion, to an industrial process. The aim: the construction, before the next decade is out more swiftly than is feasible for other fourth-generation systems of an industrial prototype, coupled to such a process. The Areva Group takes up this approach, with its ANTARES project. For the purposes of characterizing the thermal properties of heterogeneous, multi-scale materials, as a function of temperature, experimental and numerical instruments have been developed at the Microstructure and Behavior Laboratory (Laboratoire microstructure et comportement), at CEA Le Ripault Center. They have been applied to the thermal characterization of the various layers in a high temperature reactor (HTR) fuel ball. (authors)

  12. Safety philosophy of gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-01-01

    Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

  13. Corrosion behaviour of high temperature alloys in the cooling gas of high temperature reactors

    International Nuclear Information System (INIS)

    Quadakkers, W.J.; Schuster, H.

    1989-01-01

    The reactive impurities in the primary cooling helium of advanced high temperature gas cooled reactors (HTGR) can cause oxidation, carburization or decarburization of the heat exchanging metallic components. By studies of the fundamental aspects of the corrosion mechanisms it became possible to define operating conditions under which the metallic construction materials show, from the viewpoint of technical application, acceptable corrosion behaviour. By extensive test programmes with exposure times of up to 30,000 hours, a data base has been obtained which allows a reliable extrapolation of the corrosion effects up to the envisaged service lives of the heat exchanging components. (author). 6 refs, 7 figs

  14. Hydrogen Production from Nuclear Energy via High Temperature Electrolysis

    International Nuclear Information System (INIS)

    James E. O'Brien; Carl M. Stoots; J. Stephen Herring; Grant L. Hawkes

    2006-01-01

    This paper presents the technical case for high-temperature nuclear hydrogen production. A general thermodynamic analysis of hydrogen production based on high-temperature thermal water splitting processes is presented. Specific details of hydrogen production based on high-temperature electrolysis are also provided, including results of recent experiments performed at the Idaho National Laboratory. Based on these results, high-temperature electrolysis appears to be a promising technology for efficient large-scale hydrogen production

  15. High temperature applications of nuclear energy

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was organized to review industry/user needs designs, status of technology and the associated economics for high temperature applications. It was attended by approximately 100 participants from nine countries. The participants presented 17 papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  16. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  17. High-temperature reactor developments in the Netherlands

    International Nuclear Information System (INIS)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van.

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.)

  18. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.

    2017-09-26

    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  19. Structural instabilities of high temperature alloys and their use in advanced high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Schuster, H.; Ennis, P.J.; Nickel, H.; Czyrska-Filemonowicz, A.

    1989-01-01

    High-temperature, iron-nickel and nickel based alloys are the candidate heat exchanger materials for advanced high temperature gas-cooled reactors supplying process heat for coal gasification, where operation temperatures can reach 850-950 deg. C and service lives of more than 100,000 h are necessary. In the present paper, typical examples of structural changes which occur in two representative alloys (Alloy 800 H, Fe-32Ni-20Cr and Alloy 617, Ni-22Cr-12Co-9Mo-1Al) during high temperature exposure will be given and the effects on the creep rupture properties discussed. At service temperatures, precipitation of carbides occurs which has a significant effect on the creep behaviour, especially in the early stages of creep when the precipitate particles are very fine. During coarsening of the carbides, carbides at grain boundaries restrict grain boundary sliding which retards the development of creep damage. In the service environments, enhanced carbide precipitation may occur due to the ingress of carbon from the environment (carburization). Although the creep rate is not adversely affected, the ductility of the carburized material at low and intermediate temperatures is very low. During simulated service exposures, the formation of surface corrosion scales, the precipitation of carbides and the formation of internal oxides below the surface leads to depletion of the matrix in the alloying elements involved in the corrosion processes. In thin-walled tubes the depletion of Cr due to Cr 2 O 3 formation on the surface can lead to a loss of creep strength. An additional depletion effect resulting from environmental-metal reactions is the loss of carbon (decarburization) which may occur in specific environments. The compositions of the cooling gases which decarburize the material have been determined; they are to be avoided during reactor operation

  20. Neutronics of a liquid salt cooled - very high temperature reactor

    International Nuclear Information System (INIS)

    Zakova, J.

    2007-01-01

    matrix. The external radius of a TRISO particle has been set to 410μm, the radius of the fuel kernel to 150μm, in case of plutonium fueled core, and 215μm in case of uranium fueled core. The core was modelled in stochastic, three-dimensional code MCNP, version 4c3, in the finest detail. First, an under moderated core setup was found for both types of fuel by modifying the fuel to moderator ratio; then, the void and the thermal coefficients of reactivity were investigated. Few single - component molten salts were involved in the study of the void effect, in order to estimate worth of these components; NaF, BeF 2 , LiF, ZrF 4 . As a reference multi-component salt, Li 2 Be 4 F, referred to as FLiBe, was investigated. Results: It can be seen that removing BeF 2 from the core brings a negative reactivity contribution, while other three components, NaF, LiF and ZrF 4 would in a mixture contribute to the reactivity positively. Voiding FLiBe, which is a mixture of 66% of LiF and 34% BeF 2 , is equivalent to a negative reactivity insertion. Both the moderator and the fuel temperature coefficients of reactivity are large and negative for both plutonium and uranium fueled core. In the operational temperature interval (1200 K for graphite and 1500 K for fuel), the total temperature feedback is - 7.82 pcm/K for the plutonium fueled core and -2.47 pcm/K for the uranium fueled core. This results show, that the LS-VHTR core has a potential to meet the basic safety requirements as both uranium, and spent LWR fuel burner. References: [1] D. T. Ingersoll, L. J. Ott, J. P. Renier, S. J. Ball, W. R. Corwin, C. W. Forsberg, D. F. Williams, D. F. Wilson, L. Reid, G. D. Del Cul, P. F. Peterson, H. Zhao, P. S. Pickard, E. J. Parma. Status of Preconceptual Design of the Advanced High-Temperature Reactor. (ORNL, The United States of America, Tennessee 2004). [2] A. Talamo, W. Gudowski, F. Venneri, Annals of Nuclear Energy, 31, 173-196 (2004), The burnup capabilities of the Deep Burn Modular

  1. Deterministic Modeling of the High Temperature Test Reactor

    International Nuclear Information System (INIS)

    Ortensi, J.; Cogliati, J.J.; Pope, M.A.; Ferrer, R.M.; Ougouag, A.M.

    2010-01-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green's Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2-3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control

  2. Assessment of very high-temperature reactors in process applications

    International Nuclear Information System (INIS)

    Spiewak, I.; Jones, J.E. Jr.; Gambill, W.R.; Fox, E.C.

    1976-11-01

    An overview is presented of the technical and economic feasibility for the development of a very high-temperature reactor (VHTR) and associated processes. A critical evaluation of VHTR technology for process temperatures of 1400 and 2000 0 F is made. Additionally, an assessment of potential market impact is made to determine the commercial viability of the reactor system. It is concluded that VHTR process heat in the range of 1400 to 1500 0 F is attainable with near-term technology. However, process heat in excess of 1600 0 F would require considerably more materials development. The potential for the VHTR could include a major contribution to synthetic fuel, hydrogen, steel, and fertilizer production and to systems for transport and storage of high-temperature heat. A recommended development program including projected costs is presented

  3. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  4. A probabilistic consequence assessment for a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2017-02-15

    Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

  5. Design of project management system for 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Zhu Yan; Xu Yuanhui

    1998-01-01

    A framework of project management information system (MIS) for 10 MW high temperature gas-cooled test reactor is introduced. Based on it, the design of nuclear project management information system and project monitoring system (PMS) are given. Additionally, a new method of developing MIS and Decision Support System (DSS) has been tried

  6. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  7. Brazing refractory metals used in high-temperature nuclear instrumentation

    International Nuclear Information System (INIS)

    Palmer, A. J.; Woolstenhulme, C. J.

    2009-01-01

    As part of the U. S. Department of Energy (DOE)-sponsored Next Generation Nuclear Project (NGNP) currently ongoing at Idaho National Laboratory (INL), the irradiation performance of candidate high-temperature gas reactor fuels and materials is being evaluated at INL's Advanced Test Reactor (ATR). The design of the first Advanced Gas Reactor (AGR-1) TRISO fuel experiment, currently being irradiated in the ATR, required development of special techniques for brazing niobium and molybdenum. Brazing is one technique used to join refractory metals to each other and to stainless steel alloys. Although brazing processes are well established, it is difficult to braze niobium, molybdenum, and most other refractory metals because they quickly develop adherent oxides when exposed to room-temperature air. Specialized techniques and methods were developed by INL to overcome these obstacles. This paper describes the techniques developed for removing these oxides, as well as the ASME Section IX-qualified braze procedures that were developed as part of the AGR-1 project. All brazes were made using an induction coil with an inert or reducing atmosphere at low pressure. Other parameters, such as filler metals, fluxes used, and general setup procedures, are also discussed. (authors)

  8. Brazing Refractory Metals Used In High-Temperature Nuclear Instrumentation

    International Nuclear Information System (INIS)

    Palmer, A.J.; Woolstenhulme, C.J.

    2009-01-01

    As part of the U. S. Department of Energy (DOE) sponsored Next Generation Nuclear Project (NGNP) currently ongoing at Idaho National Laboratory (INL), the irradiation performance of candidate high-temperature gas reactor fuels and materials is being evaluated at INL's Advanced Test Reactor (ATR). The design of the first Advanced Gas Reactor (AGR 1) experiment, currently being irradiated in the ATR, required development of special techniques for brazing niobium and molybdenum. Brazing is one technique used to join refractory metals to each other and to stainless steel alloys. Although brazing processes are well established, it is difficult to braze niobium, molybdenum, and most other refractory metals because they quickly develop adherent oxides when exposed to room-temperature air. Specialized techniques and methods were developed by INL to overcome these obstacles. This paper describes the techniques developed for removing these oxides, as well as the ASME Section IX-qualified braze procedures that were developed as part of the AGR-1 project. All brazes were made using an induction coil with an inert or reducing atmosphere at low pressure. Other parameters, such as filler metals, fluxes used, and general setup procedures, are also discussed

  9. Brazing refractory metals used in high-temperature nuclear instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Woolstenhulme, C. J. [EG and G Services, Inc., (United States)

    2009-07-01

    As part of the U. S. Department of Energy (DOE)-sponsored Next Generation Nuclear Project (NGNP) currently ongoing at Idaho National Laboratory (INL), the irradiation performance of candidate high-temperature gas reactor fuels and materials is being evaluated at INL's Advanced Test Reactor (ATR). The design of the first Advanced Gas Reactor (AGR-1) TRISO fuel experiment, currently being irradiated in the ATR, required development of special techniques for brazing niobium and molybdenum. Brazing is one technique used to join refractory metals to each other and to stainless steel alloys. Although brazing processes are well established, it is difficult to braze niobium, molybdenum, and most other refractory metals because they quickly develop adherent oxides when exposed to room-temperature air. Specialized techniques and methods were developed by INL to overcome these obstacles. This paper describes the techniques developed for removing these oxides, as well as the ASME Section IX-qualified braze procedures that were developed as part of the AGR-1 project. All brazes were made using an induction coil with an inert or reducing atmosphere at low pressure. Other parameters, such as filler metals, fluxes used, and general setup procedures, are also discussed. (authors)

  10. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    International Nuclear Information System (INIS)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

  11. Hydrogen production from fusion reactors coupled with high temperature electrolysis

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.; Steinberg, M.

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and complement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Processes which may be considered for this purpose include electrolysis, thermochemical decomposition or thermochemical-electrochemical hybrid cycles. Preliminary studies at Brookhaven indicate that high temperature electrolysis has the highest potential efficiency for production of hydrogen from fusion. Depending on design electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

  12. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  13. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    International Nuclear Information System (INIS)

    Moreira, Uebert G.; Dominguez, Dany S.

    2017-01-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  14. Very-high-temperature gas reactor environmental impacts assessment

    International Nuclear Information System (INIS)

    Baumann, C.D.; Barton, C.J.; Compere, E.L.; Row, T.H.

    1977-08-01

    The operation of a Very High Temperature Reactor (VHTR), a slightly modified General Atomic type High Temperature Gas-Cooled Reactor (HTGR) with 1600 F primary coolant, as a source of process heat for the 1400 0 F steam-methanation reformer step in a hydrogen producing plant (via hydrogasification of coal liquids) was examined. It was found that: (a) from the viewpoint of product contamination by fission and activation products, an Intermediate Heat Exchanger (IHX) is probably not necessary; and (b) long term steam corrosion of the core support posts may require increasing their diameter (a relatively minor design adjustment). However, the hydrogen contaminant in the primary coolant which permeates the reformer may reduce steam corrosion but may produce other problems which have not as yet been resolved. An IHX in parallel with both the reformer and steam generator would solve these problems, but probably at greater cost than that of increasing the size of the core support posts. It is recommended that this corrosion problem be examined in more detail, especially by investigating the performance of current fossil fuel heated reformers in industry. Detailed safety analysis of the VHTR would be required to establish definitely whether the IHX can be eliminated. Water and hydrogen ingress into the reactor system are potential problems which can be alleviated by an IHX. These problems will require analysis, research and development within the program required for development of the VHTR

  15. Research on reactor physics using the Very High Temperature Reactor Critical Assembly (VHTRC)

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1988-01-01

    The High Temperature Engineering Test Reactor (HTTR), of which the research and development are advanced by Japan Atomic Energy Research Institute, is planned to apply for the permission of installation in fiscal year 1988, and to start the construction in the latter half of fisical year 1989. As the duty of reactor physics research, the accuracy of the nuclear data is to be confirmed, the validity of the nuclear design techniques is to be inspected, and the nuclear safety of the HTTR core design is to be verified. Therefore, by using the VHTRC, the experimental data of the reactor physics quantities are acquired, such as critical mass, the reactivity worth of simulated control rods and burnable poison rods, the temperature factor of reactivity, power distribution and so on, and the experiment and analysis are advanced. The cores built up in the VHTRC so far were three kinds having different lattice forms and degrees of uranium enrichment. The calculated critical mass was smaller by 1-5 % than the measured values. As to the power distribution and the reactivity worth of burnable poison rods, the prospect of satisfying the required accuracy for the design of the HTTR core was obtained. The experiment using a new core having axially different enrichment degree is planned. (K.I.)

  16. Safety Philosophy in Process Heat Plants Coupled to High Temperature Reactors

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2011-01-01

    With the future availability of fossil fuel resources in doubt, high temperature nuclear reactors have the potential to be an important technology in the near term. Due to a high coolant outlet temperature, high temperature reactors (HTR) can be used to drive chemical plants that directly utilize process heat. Additionally, the high temperature improves the thermodynamic efficiency of the energy utilization. Many applications of high temperature reactors exist as a thermal driving vector for endothermic chemical process plants. Hydrogen generation using the General Atomics (GA) sulfur iodine (SI) cycle is one promising application of high temperature nuclear heat. The main chemical reactions in the SI cycle are: 1. I 2 +SO 2 + 2H 2 O → 2HI + H 2 SO 4 (Bunsen reaction) 2. H 2 SO 4 → H 2 O + SO 2 + 1/2O 2 (Sulfuric acid decomposition) 3. 2HI → H 2 + I 2 (Hydrogen Iodide decomposition). With the exception of hydrogen and oxygen, all relevant reactants are recycled within the process. However, there are many unresolved safety and operational issues related to implementation of such a coupled plant

  17. Survey of high-temperature nuclear heat application

    International Nuclear Information System (INIS)

    Kirch, N.; Schaefer, M.

    1984-01-01

    Nuclear heat application at high temperatures can be divided into two areas - use of high-temperature steam up to 550 deg. C and use of high-temperature helium up to about 950 deg. C. Techniques of high-temperature steam and heat production and application are being developed in several IAEA Member States. In all these countries the use of steam for other than electricity production is still in a project definition phase. Plans are being discussed about using steam in chemical industries, oil refineries and for new synfuel producing plants. The use of nuclear generated steam for oil recovery from sands and shale is also being considered. High-temperature nuclear process heat production gives new possibilities for the application of nuclear energy - hard coals, lignites, heavy oils, fuels with problems concerning transport, handling and pollution can be converted into gaseous or liquid energy carriers with no loss of their energy contents. The main methods for this conversion are hydrogasification with hydrogen generated by nuclear heated steam reformers and steam gasification. These techniques will allow countries with large coal resources to replace an important part of their natural gas and oil consumption. Even countries with no fossil fuels can benefit from high-temperature nuclear heat - hydrogen production by thermochemical water splitting, nuclear steel making, ammonia production and the chemical heat-pipe system are examples in this direction. (author)

  18. High-temperature reactor developments in the Netherlands

    International Nuclear Information System (INIS)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    The high-temperature reactor development in the Netherlands is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTR for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will have a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. As a part of the HTR program at ECN, chemical aspects of HTR fuel and coated particles are studied. Experimental work on the oxidation resistance of coating materials and fission product attack on coating materials as well as thermochemical calculations of the fuel particles are done at ECN. The concept-HTR of ECN is fuelled with UO 2 , but the use of thorium is considered. The composition of the fuel determines the oxygen potential, which plays a key role in chemical safety of the fuel. Thermochemical calculations of the chemical form of cesium inside the HTR fuel particles were performed for a wide oxygen potential range. The chemical form of cesium determines the cesium pressure inside the fuel particle, which in turn determines the release behavior of Cs from defective particles. At normal operating temperatures and low oxygen potentials, the chemical form of cesium is C 60 Cs. It is known that cesium carbon compounds decompose above 650degC in vacuum. The stability of these compounds in the fuel particles at high temperatures(1000-1600degC) is questioned. Decomposition of these compounds may result in high cesium pressures even at normal operating conditions. Experimental work on the thermodynamic properties of cesium compounds at high temperatures is currently performed. (J.P.N.)

  19. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-01-01

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  20. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  1. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B.

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si 3 N 4 . Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

  2. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  3. High-temperature gas-cooled reactors and process heat

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

    High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

  4. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  5. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  6. The modular high-temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    Neylan, A.J.

    1986-10-01

    The MHTGR is an advanced reactor concept being developed in the USA under a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant and a graphite moderator. However the specific size and configuration are selected to utilize the inherently safe characteristics associated with these standard features coupled with passive safety systems to provide a significantly higher margin of safety and investment protection than current generation reactors. Evacuation or sheltering of the public is not required. The major components of the nuclear steam supply, with special emphasis on the core, are described. Safety assessments of the concept are discussed

  7. High temperature electrochemistry related to light water reactor corrosion

    International Nuclear Information System (INIS)

    Nagy, Gabor; Kerner, Zsolt; Balog, Janos; Schiller, Robert

    2004-01-01

    The present work deals with corrosion problems related to conditions which prevail in a WWER primary circuit. We had a two-fold aim: (A) electrochemical methods were applied to characterise the hydrothermally produced oxides of the cladding material (Zr-1%Nb) of nuclear fuel elements used in Russian made power reactors of WWER type, and (B) a number of possible reference electrodes were investigated with a view to high temperature applications. (A) Test specimens made of the cladding material, Zr-1%Nb, were immersed into an autoclave, filled with an aqueous solution typical to a WWER primary circuit, and were treated for different periods of time up to 28 weeks. The electrode potentials were measured and electrochemical impedance spectra (EIS) were taken regularly both as a function of oxidation time and temperature. This rendered information on the overall kinetics of oxide growth. By combining in situ and ex situ impedance measurements, with a particular view of the temperature dependence of EIS, we concluded that the high frequency region of impedance spectra is relevant to the presence of oxide layer on the alloy. This part of the spectra was treated in terms of a parallel CPE||R ox equivalent circuit (CPE denoting constant phase element, R ox ohmic resistor). The CPE element was understood as a dispersive resistance in terms of the continuous time random walk theory by Scher and Lax. This enabled us to tell apart electrical conductance and oxide growth with a model of charge transfer and recombination within the oxide layer as rate determining steps. (B) Three types of reference electrodes were tested within the framework of the LIRES EU5 project: (i) external Ag/AgCl, (ii) Pt/Ir alloy and (iii) Pd(Pt) double polarised active electrode. The most stable of the electrodes was found to be the Pt/Ir one. The Ag/AgCl electrode showed good stability after an initial period of some days, while substantial drifts were found for the Pd(Pt) electrode. EIS spectra of the

  8. Reactivity control system of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio

    2004-01-01

    The reactivity control system of the high temperature engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum temperature of the control rods reaches about 900 deg. C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described

  9. Design of High Temperature Reactor Vessel Using ANSYS Software

    International Nuclear Information System (INIS)

    Bandriyana; Kasmudin

    2003-01-01

    Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  11. High temperature helium-cooled fast reactor (HTHFR)

    International Nuclear Information System (INIS)

    Karam, R.A.; Blaylock, Dwayne; Burgett, Eric; Mostafa Ghiaasiaan, S.; Hertel, Nolan

    2006-01-01

    Scoping calculations have been performed for a very high temperature (1000 o C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam-UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  13. Feasibility study of high temperature reactor utilization in Czech Republic after 2025

    Energy Technology Data Exchange (ETDEWEB)

    Losa, Evžen, E-mail: evzen.losa@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Souček, Václav; Kohout, Petr [AZIN CZ, s.r.o., Hanusova 3, 140 00 Praha 4 (Czech Republic)

    2014-05-01

    High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country.

  14. Feasibility study of high temperature reactor utilization in Czech Republic after 2025

    International Nuclear Information System (INIS)

    Losa, Evžen; Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír; Souček, Václav; Kohout, Petr

    2014-01-01

    High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country

  15. A high temperature reactor could be used to eliminate the Russian military plutonium

    International Nuclear Information System (INIS)

    Foucher, N.

    1999-01-01

    The GT-MHR reactor (Gas Turbine Modular Helium Reactor) aims the double objective to eliminate the Russian plutonium coming from weapons, ( until 3 tons by year) and to produce a competitive energy from a small-scale power reactor with a nuclear fuel that can be of different type (plutonium or uranium). This reactor has several advantages: a high yield (47%) as every high temperature reactor and to be used in combined cycle, a high level of safety because of its ability to evacuate the residual power in a totally passive way and because of the nature of its fuel that is made of ceramics with a very high melting point that is to say no possibility of core melt. The fission products are contained in the ceramics so that reactor cannot disseminate radioactivity in its structure and consequently does not induce irradiation for the personnel. (N.C.)

  16. Fatigue crack growth characteristics of a533 brade b glass i plate in an environment of high-temperature primary grade nuclear reactor water

    International Nuclear Information System (INIS)

    Mager, T.R.; Moon, D.M.; Landes, J.D.

    1976-01-01

    To characterize the effect of environment on crack growth rate properties of reactor pressure vessel materials, a program was initiated as part of the Heavy Section Steel Technology Program (HSST) to evaluate the effect of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) primary grade coolant environments. The experiments included such variables as frequency, temperature and R ratio. This paper describes the investigation and presents the results of a fracture mechanics evaluation of the fatigue crack growth rate tests of A533 Grade B Class 1 steel plate material in an environment of primary reactor grade water at 550 0 F (288 0 C). A compliance crack growth monitoring technique was utilized to measure the crack growth. The compliance crack length monitor uses a linear variable differential transformer (LVDT) to measure the specimen front face displacement which is converted to crack length by the appropriate compliance calibration curve. The crack growth rate tests were conducted on constant load universal fatigue machines, under sinusoidal tension to tension loading conditions. Tests showed an increase in growth rates at a frequency of 1 cpm over previous results obtained at frequencies of 60 cpm and higher. This increase, the general character of the crack growth rate versus the $DELTA$K curve, and the results from fractographic studies, all indicated that stress corrosion cracking might have occurred for this material and environment. However, a specimen loaded statically in a PWR environment showed no static load crack growth. 13 refs

  17. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  18. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    International Nuclear Information System (INIS)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  19. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

    2005-01-01

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  20. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  1. ARCHER Project: Progress on Material and component activities for the Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.

    2014-01-01

    The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) integrated project is a four year project which was started in 2011 as part of the European Commission 7th Framework Programme (FP7) to perform High Temperature Reactor technology R&D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research & Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on ARCHER materials and component activities since the start of the project and underlines some of the main conclusions reached. (author)

  2. Material and component progress within ARCHER for advanced high temperature reactor

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

    2014-01-01

    The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

  3. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  4. Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

    International Nuclear Information System (INIS)

    Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

    2007-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered

  5. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  6. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  7. Incoloy 800 stands up to radiation and corrosion in high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Incoloy 800 has been selected for heat exchangers in helium cooled nuclear reactor prototypes for exposure to 350 to 800 0 C helium and high temperature high purity water and steam. 304H stainless steel used in heat exchangers in original design cracked in the superheater area, bellows and tubing after static pressure tests but before exposure to steam. Residual stress, chlorides, and oxygen were deduced to have caused the failures

  8. Use of nuclear process heat from pebble bed high temperature reactors to obtain oil by tertiary recovery methods. Final report. Vol. 2

    International Nuclear Information System (INIS)

    Jaeger, H.; Hermges, H.; Kammel, R.; Kugeler, K.; Phlippen, P.W.; Scheuch, H.; Schmidt, F.; Schmidtlein, P.; Schreiner, P.

    1988-01-01

    Volume II of this report examines: 1. The use of nuclear process heat in the further processing of crude oil (refinery processes, heat coupling, steam reforming helium-heated pipe furnace), 2. Analyses for process questions, 3. Questions of economy, 4. Environmental aspects, 5. Work on underground methanisation. (RB) [de

  9. Hydrogen production by water-splitting using heat supplied by a high-temperature reactor

    International Nuclear Information System (INIS)

    Courvoisier, P.; Rastoin, J.; Titiliette, Z.

    1976-01-01

    Some aspects of the use of heat of nuclear origin for the production of hydrogen by water-splitting are considered. General notions pertaining to the yield of chemical cycles are discussed and the heat balance corresponding to two specific processes is evaluated. The possibilities of high temperature reactors, with respect to the coolant temperature levels, are examined from the standpoint of core design and technology of some components. Furthermore, subject to a judicious selection of their characteristics, these reactors can lead to excellent use of nuclear fuel. The coupling of the nuclear reactor with the chemical plant by means of a secondary helium circuit gives rise to the design of an intermediate heat exchanger, which is an important component of the overall installation. (orig.) [de

  10. The use and development of the high-temperature reactor (HTR) in China. A conference report

    International Nuclear Information System (INIS)

    Marnet, C.

    2001-01-01

    Gas-cooled graphite-moderated reactors have been under development since the early days of nuclear technology. Starting with plants in Britain and France, reactors employing this combination of coolant and moderator were used in commercial nuclear power plants in the second half of the fifties. At the same time, efforts seeking to use inert helium gas as a coolant resulted in the construction in several countries, the United States and Germany in particular, of larger nuclear power plants with higher coolant temperatures and the resultant thermodynamic advantages of high efficiencies and the option of process heat generation. Economic and political considerations led to the decommissioning of these plants. Today, research and development of high-temperature reactors are concentrated on smaller units. Work is carried out in close international cooperation, especially on project designs and newly commissioned plants in China (HTR-10), Japan (HTTR, Oarai), And South Africa (ESKOM project), but also in the USA and in Russia. (orig.) [de

  11. Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

    Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

  12. First conceptual design of the experimental multi-purpose high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsunoda, T [Fuji Electric Co. Ltd., Tokyo (Japan)

    1976-02-01

    A part of the multi-purpose high temperature reactor (VHTR) was designed by the First Atomic Power Industry Group (FAPIG). Both Fuji Electric Co., Ltd. and Kawasaki Heavy Industries, Ltd. of the FAPIG group took charge of the design of main parts of the reactor Kobe Steel, Ltd., Ebara Manufacturing Co., Ltd., Shimizu Construction Co., Ltd. and the Nuclear Fuel Corp. have associated with this group. The reactor system includes a nuclear reactor and two cooling loops provided through intermediate heat exchangers in order to utilize the heat of helium gas delivered from the reactor outlet at 1,000 deg C. One is a reformer loop to produce the reducing gas for steel manufacture. The other is a testing loop for a reducing gas heater and a gas turbine. These loops transfer heat of about 25 MW at 930 deg C at rated capacity. The reformer can supply the reducing gas equivalent to the production of 100 tons per day sponge iron. A housing of the reactor is composed of a primary steel container, internal concrete and a secondary container made of reinforced concrete. The construction is based on the following principles. (1) For the very high temperature portion at 1,000 deg C, a non-metallic material such as graphite should be used. (2) The metallic construction shall be cooled with return gas below 400 deg C. (3) The steel pressure vessel shall be employed. (4) The design shall be based on the existing gas furnace.

  13. A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

    Science.gov (United States)

    Syarip; Po, L. C. C.

    2018-05-01

    In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

  14. The market potential of the high-temperature reactor

    International Nuclear Information System (INIS)

    Knizia, K.

    1982-01-01

    It seems urgently necessary to plant the power supplies of the industrial countries in a more flexible fashion and thereby to make them more susceptible to disturbances arising externally through restrictions on fuel deliveries or price increases. The development and introduction of advanced reactors and the further development of coal processing are of prime importance in this respect. This development will lead in the long-term to the production of hydrogen with the use of nuclear energy. (orig.) [de

  15. Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

    International Nuclear Information System (INIS)

    Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

    2005-01-01

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

  16. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    Bende, E.E.

    1997-01-01

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k ∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GW e a is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GW e a, respectively. (author)

  17. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

    Lockett, G.E.; Kinkead, A.N.

    1967-02-01

    In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

  18. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

    It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

  19. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  20. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    International Nuclear Information System (INIS)

    Ingersoll, D.T.

    2004-01-01

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  1. Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2003-01-01

    On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  3. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  4. Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

  5. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  6. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  7. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  8. Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.

    1976-12-01

    In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

  9. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing of high-temperature gas-cooled reactor fuel containing U-233 and thorium

    International Nuclear Information System (INIS)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed

  10. The world trends of high temperature gas-cooled reactors and the mode of utilization

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi; Shimokawa, Jun-ichi

    1974-01-01

    After a long period of research and development, high temperature gas-cooled reactors are going to enter the practical stage. The combination of a HTGR with a closed cycle helium gas turbine is advantageous in thermal efficiency, reduction of environmental impact and economy. In recent years, the direct utilization of nuclear heat energy in industries has been attracting interest. The multi-purpose utilization of high temperature gas-cooled reactors is thus now the world trend. Reviewing the world developments in this field, the following matters are described: (1) development of HTGRs in the U.K., West Germany, France and the United States; (2) development of He gas turbine, etc. in West Germany; and (3) multi-purpose utilization of HTGRs in West Germany and Japan. (Mori, K.)

  11. Analysis of a computational benchmark for a high-temperature reactor using SCALE

    International Nuclear Information System (INIS)

    Goluoglu, S.

    2006-01-01

    Several proposed advanced reactor concepts require methods to address effects of double heterogeneity. In doubly heterogeneous systems, heterogeneous fuel particles in a moderator matrix form the fuel region of the fuel element and thus constitute the first level of heterogeneity. Fuel elements themselves are also heterogeneous with fuel and moderator or reflector regions, forming the second level of heterogeneity. The fuel elements may also form regular or irregular lattices. A five-phase computational benchmark for a high-temperature reactor (HTR) fuelled with uranium or reactor-grade plutonium has been defined by the Organization for Economic Cooperation and Development, Nuclear Energy Agency (OECD NEA), Nuclear Science Committee, Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles. This paper summarizes the analysis results using the latest SCALE code system (to be released in CY 2006 as SCALE 5.1). (authors)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  13. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  14. Is tritium an issue for high temperature reactors?

    Energy Technology Data Exchange (ETDEWEB)

    Fütterer, Michael A., E-mail: michael.fuetterer@ec.europa.eu [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); D’Agata, Elio [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Raepsaet, Xavier [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DM2S, 91191 Gif-sur-Yvette Cedex (France)

    2016-09-15

    In a high temperature reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy.

  15. Is Tritium an Issue for High Temperature Reactors?

    International Nuclear Information System (INIS)

    Fütterer, Michael A.; D'Agata, Elio; Raepsaet, Xavier

    2014-01-01

    In a High Temperature Reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy. (author)

  16. High temperature reactor for the production of low temperature heat

    International Nuclear Information System (INIS)

    Muehlensiep, J.

    1986-12-01

    In this report the conditions of nuclear working reactors for district heating are described for the use in suburban areas. The design of a HTR is analysed under the point of view of safety and costs for the components and for the arrangement possibilities. The size of system is chosen by analysing important parameters for construction. The layout is determined by the retention of fission products in the coated particles of the fuel under conditions of hypothetical accidents. Based on stated data a HTR reactor for district heating will be designed. The speciality is a square shaped core which has the advantage to conduct the afterheat fastly to the outside of the pressure vessel in case of hypothetical accidents. Caused by the shape of the core the heat exchangers may be installed next to the core, the shutdown rods are maintained into reflector borings where they have a high efficiency. The whole primary circuit is surrounded by the reactor pressure vessel and is adjusted in an underground concrete cell. (orig./GL) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  18. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  19. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Bjornard, Trond; Hockert, John

    2011-01-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

  20. Graphite materials for nuclear reactors

    International Nuclear Information System (INIS)

    Oku, Tatsuo

    1991-01-01

    Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

  1. Present status and prospects of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1995-01-01

    It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors. Hence, efforts are to be continuously devoted to establish and upgrade High Temperature Gas-cooled Reactor (HTGR) technologies and to make much of research resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950degC at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). The construction of the HTTR started in March 1991, with first criticality in 1998 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been nearly completed and its main components, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas pipings and core support structures, have been manufactured at their factories and delivered to the Oarai Research Establishment of the JAERI for their installation in the middle of 1994. Fuel fabrication will be started as well. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. The IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR, such as steam reforming of methane and thermochemical water splitting for hydrogen production, was launched successfully in January 1994. Some heat utilization system is planned to be connected to the HTTR and demonstrated at the former stage of the second core. At present, steam-reforming of methane is the first candidate. The JAERI also plans to conduct material

  2. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  4. Basic design and economical evaluation of Gas Turbine High Temperature Reactor 300 (GTHTR300)

    International Nuclear Information System (INIS)

    Kazuhiko, Kunitomi; Shusaku, Shiozawa; Xing, Yan

    2007-01-01

    High Temperature Gas-cooled Reactor (HTGR) combined with a direct cycle gas turbine offers one of the most promising nuclear electricity generation options after 2010. Japan Atomic Energy Agency has been engaging in the basic design and development of Gas Turbine High Temperature Reactor 300 (GTHTR300) since 2003. Costs of capital, fuel, and operation and maintenance have been estimated. The capital cost of the GTHTR300 is lower than that of the existing light water reactor (LWR) because the generation efficiency is considerably higher whereas the construction cost is lower owing to the design simplicity of the gas turbine power conversion unit and the reactor safety system. The fuel cost is shown to equal that of LWR. The operation and maintenance cost has a slight advantage due to the use of chemically inert helium coolant. In sum, the cost of electricity for the GTHTR300 is estimated to be below US 3.3 cents/kWh (4 yen/kWh), which is about two-third of that of current LWRs in Japan. The results confirm that the net power generation cost of the GTHTR300 is much lower than that of the LWR, indicating that the GTHTR300 plant consisting of small-scale reactor units can be economically competitive to the latest large-scale LWR. (authors)

  5. Preliminary risk analysis of an Hydrogen production plant using the reformed process of methane with vapor coupled to a high temperature nuclear reactor; Analisis preliminar de riesgo de una planta de produccion de hidrogeno utilizando el proceso de reformado de metano con vapor acoplada a un reactor nuclear de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Flores y Flores, A. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. E-mail: alain_fyf@yahoo.com; Nelson E, P.F.; Francois L, J.L. [Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2004-07-01

    It is necessary to identify the different types of dangers, as well as their causes, probabilities and consequences of the same ones, inside plants, industries and any process to classify the risks. This work is focused in particular to a study using the technical HAZOP (Hazard and Operability) for a plant of reformed of methane with vapor coupled to a nuclear reactor of the type HTTR (High Temperature Test Reactor), which is designed to be built in Japan. In particular in this study the interaction is analyzed between the nuclear reactor and the plant of reformed of methane with vapor. After knowing the possible causes of risk one it is built chart of results of HAZOP to have a better vision of the consequences of this faults toward the buildings and constructions, to people and the influence of the fault on each plant; for what there are proposed solutions to mitigate these consequences or to avoid them. The work is divided in three sections: a brief introduction about the technique of HAZOP; some important aspects of the plant of reformed of methane with vapor; and the construction of the chart of results of HAZOP. (Author)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  8. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

  9. Flaw assessment procedure for high temperature reactor components

    International Nuclear Information System (INIS)

    Ainsworth, R.A.; Takahashi, Y.

    1990-01-01

    An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the USA, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program. 20 refs

  10. Recommendations for the development of high temperature reactors (HTRs)

    International Nuclear Information System (INIS)

    Bastien, D.; Chevalier, A.; Haverkate, B.R.W.; Hittner, D.; Magill, J.; Proto, G.; Reutler, H.; Taylor, N.; Lensa, W. von

    2001-01-01

    The Concerted Action ''Assessment of safety and innovative technology for an HTR power generating plant (INNOHTR)'' is funded by the European Commission within its 4th RTD Framework Programme. It is a joint project of four European nuclear industry organisations and five European research centres. This paper presents the preliminary conclusions of the INNOHTR Concerted Action. It addresses existing experience in designing, building, and operating HTRs in the areas of fuel, reactor core, and plant systems. It also addresses the need for additional research and development to support forthcoming industrial prototypes as well as for long term advanced system development. Specific areas of development are recommended along with the need for international cooperation. (author)

  11. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

    The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

  12. Comparative Analysis of Single and Dual Irradiation Pass of Deep Burn High Temperature Reactor Scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Jo, Chang Keun; Noh, Jae Man

    2012-01-01

    A concept of a deep-burn (DB) of trans uranic (TRU) elements in a high temperature reactor (HTR) has been proposed and studied with a single irradiation pass. However, there is still a significant amount of TRU after burn in an HTR. Therefore, it is necessary to burn more TRU in a fast reactor (FR) with repeated reprocessing such as a pyro-process. In this study, the fuel cycle calculations are performed and the results are compared for a singlepass DB-HHR scenario and a dual-pass sodium-cooled fast reactor (SFR) scenario. For the analysis, front-end and back-end parameters are compared. The calculations were performed by the DANESS (Dynamic Analysis of Nuclear Energy System Strategies), which is an integrated system dynamic fuel cycle analysis code

  13. Conceptual designs for advanced, high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

    2000-07-01

    AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

  14. Consideration of emergency source terms for pebble-bed high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tao, Liu; Jun, Zhao; Jiejuan, Tong; Jianzhu, Cao

    2009-01-01

    Being the last barrier in the nuclear power plant defense-in-depth strategy, emergency planning (EP) is an integrated project. One of the key elements in this process is emergency source terms selection. Emergency Source terms for light water reactor (LWR) nuclear power plant (NPP) have been introduced in many technical documents, and advanced NPP emergency planning is attracting attention recently. Commercial practices of advanced NPP are undergoing in the world, pebble-bed high-temperature gas-cooled reactor (HTGR) power plant is under construction in China which is considered as a representative of advanced NPP. The paper tries to find some pieces of suggestion from our investigation. The discussion of advanced NPP EP will be summarized first, and then the characteristics of pebble-bed HTGR relating to EP will be described. Finally, PSA insights on emergency source terms selection and current pebble-bed HTGR emergency source terms suggestions are proposed

  15. Critical evaluation of high-temperature gas-cooled reactors applicable to coal conversion

    International Nuclear Information System (INIS)

    Spiewak, I.; Jones, J.E. Jr.; Rittenhouse, P.L.; DeStefano, J.R.; Delene, J.G.

    1975-12-01

    A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 2000 0 F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 1500 0 F could be developed with a high degree of assurance. Process heat at 1600 0 F would require considerably more materials development. While temperatures up to 2000 0 F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR

  16. Conceptual designs for very high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    2000-07-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

  17. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

    2000-01-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

  18. The modular high-temperature gas-cooled reactor (MHTGR) in the US

    International Nuclear Information System (INIS)

    Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

    1987-01-01

    GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

  19. Evaluation of high temperature gas reactor for demanding cogeneration load follow

    International Nuclear Information System (INIS)

    Yan, Xing L.; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Hino, Ryutaro

    2012-01-01

    Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEA's evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEA's operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900degC heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity. (author)

  20. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  1. The production of refined intermediate fuels with high temperature reactors

    International Nuclear Information System (INIS)

    Nowacki, P.J.

    1977-01-01

    Power plants can be divided into conventional steam plants, fueled with hard coal, lignite or liquid fuel, hydroelectric plants and nuclear plants, their chief use was or is the production of electric energy and - in certain cases only - of production of process heat, using steam or hot water for process heat in industry and district heating for residential and commercial purposes. The part played by electricity in the whole energy demand is of the order of 10% to 25% the total demand, the rest is necessary for supplying process heat below 200 0 C or above 200 0 C, up to some 1500 0 C. The present distribution of energy demands is covered chiefly by liquid fuel, coal and lignite, water energy and increasing steps by nuclear fuel. It is well known that the erection of nuclear energy plants is a necessity for today and for the future. There is another necessity, i.e. to utilize the primary energy resources in a complex way i.e. to supply electricity as energy vector and other fuels as process heat as new energy vectors. These manmade fuels - whether in a gaseous or liquid phase - contain hydrogen, and one can believe, the world is entering a new energy civilisation in utilizing hydrogen and its compounds as second energy vector. The author has taken up the task to investigate this new problem of process, heat in the form of hydrogen and its compounds, by evaluating their present and future production, based on the utilization of natural gas, oil coal, water and the nuclear heat of helium, available in a closed circuit as primary coolant in a High - Temeprature Helium cooled reactor, which is symbolized in the paper as HTR. The paper deals in more detail with the following application of Nuclear Heat: hydrogasification, direct reduction of ore, mainly iron ores, ammonia synthesis, methanol synthesis Hydrocracking, long distance transfer of process heat (chemical heat pipe), hydrogenation of coal, Fischer - Tropsch synthesis, oxosynthesis, coal gasification, coal

  2. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  3. Renewable side reflector structure for a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Martin, Roger.

    1977-01-01

    The description is given of a renewable side reflector structure for a pebble bed high temperature reactor of the kind comprising a cylindrical graphite vessel constituting the neutron reflector, this vessel being filled with graphite pebbles containing the nuclear fuel and enclosed in a concrete protective containment. The internal peripheral area of the vessel is constituted by a line of adjacent graphite rods mounted so that they can rotate about their longitudinal axis and manoeuvrable from outside the concrete containment by means of a shaft passing into it [fr

  4. Study on fundamental features of helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2004-01-01

    The High temperature gas-cooled reactor (HTGR) coupled with helium turbine cycle is considered as one of the leading candidates for future nuclear power plants. The HTGR helium turbine cycle was analyzed and optimized. Then the focal point of investigation was concentrated on the fundamental thermodynamic and aerodynamic features of helium turbomachine. As a result, a helium turbomachine is different from a general combustion gas turbine in two main design features, that is a helium turbomachine has more blade stages and shorter blade length, which are caused by the helium property and the high pressure of a closed cycle, respectively. (authors)

  5. A prestressed concrete pressure vessel for helium high temperature reactor system

    International Nuclear Information System (INIS)

    Horner, R.M.W.; Hodzic, A.

    1976-01-01

    A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

  6. Optimizing a High-Temperature Hydrogen Co-generation Reactor for Both Economic and Environmental Performance

    International Nuclear Information System (INIS)

    Weimar, Mark R.; Wood, Thomas W.; Reichmuth, Barbara A.; Johnson, Wayne L.

    2003-01-01

    This paper analyzes outcomes for a 3000 MWt High Temperature Gas Reaction nuclear power plant, given price and cost assumptions, and determined what level of hydrogen and electricity production would optimize the plant economically and environmentally (carbon reduction). The tradeoff between producing hydrogen through steam methane reformation and producing electricity is so disproportionate, that advanced reactors will likely be used only as peaking plants for electricity unless policymakers intervene with incentives to change the mix of electricity and hydrogen. The magnitude of the increase in electric prices or decrease in hydrogen prices required to allow electricity production indicate that substantial error in cost estimates would be required to change our analysis.

  7. Resource utilization of symbiotic high-temperature gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; Brogli, R.H.

    1978-01-01

    The cumulative uranium requirements of different symbiotic combinations of high-temperature gas-cooled reactor (HTGR) prebreeders have been calculated assuming an open-end nuclear economy. The results obtained indicate that the combination of prebreeders and near-breeders does not save resources over a self-generated recycle case of comparable conversion ratio, and that it may take between 40 and 50 yr before the symbiotic system containing breeders starts saving resources over an HTGR with self-generated recycle and a conversion ratio of 0.83

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  10. High temperature alloys for the primary circuit of a prototype nuclear process heat plant

    International Nuclear Information System (INIS)

    Ennis, P.J.; Schuster, H.

    1979-01-01

    As part of a comprehensive materials test programme for the High Temperature Reactor Project 'Prototype Plant for Nuclear Process Heat' (PNP), high temperature alloys are being investigated for primary circuit components operating at temperatures above 750 0 C. On the basis of important material parameters, in particular corrosion behaviour and mechanical properties in primary coolant helium, the potential of candidate alloys is discussed. By comparing specific PNP materials data with the requirements of PNP and those of conventional plant, the implications for the materials programme and component design are given. (orig.)

  11. Moderator for nuclear reactor

    International Nuclear Information System (INIS)

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  12. EUROPAIRS: The European project on coupling of High Temperature Reactors with industrial processes

    International Nuclear Information System (INIS)

    Angulo, C.; Bogusch, E.; Bredimas, A.; Delannay, N.; Viala, C.; Ruer, J.; Muguerra, Ph.; Sibaud, E.; Chauvet, V.; Hittner, D.; Fütterer, M.A.; Groot, S. de; Lensa, W. von; Verfondern, K.; Moron, R.; Baudrand, O.; Griffay, G.; Baaten, A.; Segurado-Gimenez, J.

    2012-01-01

    Developers of High Temperature Reactors (HTR) worldwide acknowledge that the main asset for market breakthrough is its unique ability to address growing needs for industrial cogeneration of heat and power (CHP) owing to its high operating temperature and flexibility, adapted power level, modularity and robust safety features. A strong alliance between nuclear and process heat user industries is a necessity for developing such a nuclear system for the conventional process heat market, just as the electro-nuclear development required a close partnership with utilities. Initiating such an alliance is one of the objectives of the EUROPAIRS project ( (www.europairs.eu)) presently on-going in the frame of the Euratom 7th Framework Programme (FP7). Although small and of short duration (21 months), EUROPAIRS is of strategic importance: it generates the boundary conditions for rapid demonstration of collocating HTR with industrial processes as proposed by the European High Temperature Reactor Technology Network (HTR-TN). This paper presents the main goals, the organization and the working approach of EUROPAIRS. It also presents the status of the viability assessment studies for coupling HTR with industrial end-user systems as one of the main pillars of the project. The main goal of the viability assessment is to identify developments required to remove the last technological and licensing barriers for a viable coupling scheme. The study is expected to result in guidelines for directing the choice of an industrial scale prototype.

  13. EUROPAIRS: The European project on coupling of High Temperature Reactors with industrial processes

    Energy Technology Data Exchange (ETDEWEB)

    Angulo, C., E-mail: carmen.angulo@gdfsuez.com [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium); Bogusch, E. [AREVA NP GmbH, Paul-Gossen-Strasse 100, 91052 Erlangen (Germany); Bredimas, A. [LGI Consulting, 37 rue de la Grange aux Belles, 75010 Paris (France); Delannay, N. [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium); Viala, C. [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France); Ruer, J.; Muguerra, Ph.; Sibaud, E. [SAIPEM S.A., 1/7 Avenue San Fernando, 78884 Saint Quentin en Yvelines Cedex (France); Chauvet, V. [LGI Consulting, 37 rue de la Grange aux Belles, 75010 Paris (France); Hittner, D. [AREVA NP Inc., 3315 Old Forest Road, Lynchburg, VA 24501 (United States); Fuetterer, M.A. [European Commission, Joint Research Centre, 1755ZG Petten (Netherlands); Groot, S. de [Nuclear Research and Consultancy Group, 1755ZG Petten (Netherlands); Lensa, W. von; Verfondern, K. [Forschungszentrum Juelich GmbH, Leo-Brandt-Strasse,52425 Juelich (Germany); Moron, R. [Solvay SA, rue du Prince Albert 33, 1050 Brussels (Belgium); Baudrand, O. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP 17, 92262 Fontenay-aux-Roses cedex (France); Griffay, G. [Arcelor Mittal Maizieres Research SA, rue Luigi Cherubini 1A5, 39200 Saint Denis (France); Baaten, A. [USG/Baaten Energy Consulting, Burgermeester-Ceulen-Straat 78, 6212CT Maastricht (Netherlands); Segurado-Gimenez, J. [Tractebel Engineering S.A. (GDF SUEZ), Avenue Ariane 7, 1200 Brussels (Belgium)

    2012-10-15

    Developers of High Temperature Reactors (HTR) worldwide acknowledge that the main asset for market breakthrough is its unique ability to address growing needs for industrial cogeneration of heat and power (CHP) owing to its high operating temperature and flexibility, adapted power level, modularity and robust safety features. A strong alliance between nuclear and process heat user industries is a necessity for developing such a nuclear system for the conventional process heat market, just as the electro-nuclear development required a close partnership with utilities. Initiating such an alliance is one of the objectives of the EUROPAIRS project ( (www.europairs.eu)) presently on-going in the frame of the Euratom 7th Framework Programme (FP7). Although small and of short duration (21 months), EUROPAIRS is of strategic importance: it generates the boundary conditions for rapid demonstration of collocating HTR with industrial processes as proposed by the European High Temperature Reactor Technology Network (HTR-TN). This paper presents the main goals, the organization and the working approach of EUROPAIRS. It also presents the status of the viability assessment studies for coupling HTR with industrial end-user systems as one of the main pillars of the project. The main goal of the viability assessment is to identify developments required to remove the last technological and licensing barriers for a viable coupling scheme. The study is expected to result in guidelines for directing the choice of an industrial scale prototype.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  15. Nuclear process heat at high temperature: Application, realization and development programme

    International Nuclear Information System (INIS)

    Sammeck, K.H.; Fischer, R.

    1976-01-01

    Studies in the Federal Republic of Germany (FRG), the USA and the United Kingdom have shown that high-temperature helium energy from an HTR can advantageously be utilized for coal gasification and other fossil fuel conversion processes, and that a substantial demand for substitute natural gas (SNG) can be expected in the future. These results are based on plant design studies, economic assessments and basic development efforts in the field of coal gasification with nuclear heat, which in the FRG were carried out by Arbeitsgemeinschaft Nukleare Prozesswaerme (ANP)-members, HRB and KFA Juelich. Nuclear process plants are based on different gasification processes, resulting in different concepts of the nuclear heat system. In the case of hydro-gasification it is expected that steam reformers, arranged within the primary circuit of the reactor, will be heated directly by the primary helium. In the case of steam gasification, the high-temperature energy must be transferred to the gasification process via an intermediate circuit which is coupled to a gasifier outside the containment. In both cases the design of the nuclear reactor resembles an HTR for electricity generation. The main objectives of the development of nuclear process heat are to increase the helium outlet temperature of the reactor up to 950 0 C, to develop metallic alloys for high-temperature components such as heat exchangers, to design and construct a hot-gas duct, a steam reformer and a helium-helium heat exchanger and to develop the gasification processes. The nuclear safety regulations and the interface problems between the reactor, the process plant and the electricity generating plant have to be considered thoroughly. The Arbeitsgemeinschaft Nukleare Prozesswaerme and HRB started a development programme, in close collaboration with KFA Juelich, which will lead to the construction of a prototype plant for coal gasification with nuclear heat within 5 to 5 1/2 years. A survey of the main objectives

  16. Research and development for high temperature gas cooled reactor in Japan

    International Nuclear Information System (INIS)

    Taketani, K.

    1978-01-01

    The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

  17. Preliminary study on helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Chen Yihua; Wang Jie; Zhang Zuoyi

    2003-01-01

    In the high temperature gas-cooled reactor (HTGR), gas turbine cycle is a new concept in the field of nuclear power. It combines two technologies of HTGR and gas turbine cycle, which represent the state-of-the-art technologies of nuclear power and fossil fuel generation respectively. This approach is expected to improve safety and economy of nuclear power plant significantly. So it is a potential scheme with competitiveness. The heat-recuperated cycle is the main stream of gas turbine cycle. In this cycle, the work medium is helium, which is very different from the air, so that the design features of the helium turbomachine and combustion gas turbomachine are different. The paper shows the basic design consideration for the heat-recuperated cycle as well as helium turbomachine and highlights its main design features compared with combustion gas turbomachine

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  20. Evaluation of the security of a hydrogen producing plant by means of the S I cycle coupled to a nuclear reactor of high temperature; Evaluacion de la seguridad de una planta productora de hidrogeno mediante el ciclo SI acoplada a un reactor nuclear de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz S, T.; Francois, J. L.; Nelson, P. F. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Cruz G, M. J., E-mail: truizsmx@yahoo.com.mx [UNAM, Facultad de Quimica, Ciudad Universitaria, 04510 Mexico D. F. (MX)

    2011-11-15

    At the present one of the processes that demonstrates, theoretically, to be one of the most efficient for the hydrogen production is the thermal-chemistry cycle Sulfur-Iodine. One way of obtaining the temperature ranges required by the process is through the helium coming from a very high temperature reactor. The coupling of the chemical plant with the nuclear plant presents aspects of security that should be analyzed; among them the analysis of the danger of the process materials is, with the purpose of implementing security measures to protect the facilities and equipment s, the environment and the population. These measures can be: emergency answer plans of the stations, definition of the minimum distance required among facilities, determination of the exclusion area, etc. In this study simulations were made with the computer code Phast in order to knowing the possible affectation areas due to the liberation of a great quantity of energy due to a helium leak to very high temperature, of toxic materials or by a possible hydrogen combustion. The results for the liberations of sulfuric acid, hydrogen, iodine, helium and sulfur dioxide are shown, specially. The operation conditions were taken of a combination of the preliminary design proposed by General Atomics and the optimized conditions by the Korea Advanced Institute of Science and Technology, considering a production of 1 kg-mol/s of hydrogen. The iodine was the material that presented a major affectation area. (Author)

  1. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  2. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  3. Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

  4. Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    2003-08-01

    Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

  5. Assessment of very high-temperature reactors in process applications. Appendix III. Engineering evaluation of process heat applications for very-high temperature reactors

    International Nuclear Information System (INIS)

    Wiggins, D.S.; Williams, J.J.

    1977-04-01

    An engineering and economic evaluation is made of coal conversion processes that can be coupled to a very high-temperature nuclear reactor heat source. The basic system developed by General Atomic/Stone and Webster (GA/S and W) is similar to the H-coal process developed by Hydrocarbon Research, Inc., but is modified to accommodate a nuclear heat source and to produce synthetic natural gas (SNG), synthesis gas, and hydrogen in addition to synthetic crude liquids. The synthetic crude liquid production is analyzed by using the GA/S and W process coupled to either a nuclear- or fossil-heat source. Four other processes are included for comparison: (1) the Lurgi process for production of SNG, (2) the Koppers-Totzek process for production of either hydrogen or synthesis gas, (3) the Hygas process for production of SNG, and (4) the Westinghouse thermal-chemical water splitting process for production of hydrogen. The production of methanol and iron ore reduction are evaluated as two potential applications of synthesis gas from either the GA/S and W or Koppers-Totzek processes. The results indicate that the product costs for each of the gasification and liquefaction processes did not differ significantly, with the exception that the unproven Hygas process was cheaper and the Westinghouse process considerably more expensive than the others

  6. Production of energy in a high temperature reactor

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The cooling gas having left the reactor core is fed to a generator for direct production of current from the kinetic energy. Afterwards the gas is fed to a heat exchanger for cooling, then compressed and refed to the reactor core. The method further comprises that one part of the energy of the fission material is directly converted to electric energy in the reactor core, whereas the other part of the energy of the fission material is impressed upon the cooling gas. According to the invention the cooling gas when entering the reactor is first fed to that part of the reactor core which serves as a thermoionic or thermoelectric transducer. Afterwards the cooling gas is fed to the remaining part of the reactor gas. (P.K.)

  7. Research and development program of hydrogen production system with high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Miyamoto, Y.; Shiozawa, S.; Ogawa, M.; Inagaki, Y.; Nishihara, T.; Shimizu, S.

    2000-01-01

    Japan Atomic Energy Research Institute (JAERI) has been developing a hydrogen production system with a high temperature gas-cooled reactor (HTGR). While the HTGR hydrogen production system has the following advantages compared with a fossil-fired hydrogen production system; low operation cost (economical fuel cost), low CO 2 emission and saving of fossil fuel by use of nuclear heat, it requires some items to be solved as follows; cost reduction of facility such as a reactor, coolant circulation system and so on, development of control and safety technologies. As for the control and safety technologies, JAERI plans demonstration test with hydrogen production system by steam reforming of methane coupling to 30 Wt HTGR, named high temperature engineering test reactor (HTTR). Prior to the demonstration test, a 1/30-scale out-of-pile test facility is in construction for safety review and detailed design of the HTTR hydrogen production system. Also, design study will start for reduction of facility cost. Moreover, basic study on hydrogen production process without CO 2 emission is in progress by thermochemical water splitting. (orig.)

  8. Basic study on high temperature gas cooled reactor technology for hydrogen production

    International Nuclear Information System (INIS)

    Chang, Jong Hwa; Lee, W. J.; Lee, H. M.

    2003-01-01

    The annual production of hydrogen in the world is about 500 billion m 3 . Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

  9. Operating experience with the DRAGON High Temperature Reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.A.; Capp, P.D.

    2002-01-01

    The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

  10. Failure mechanisms in high temperature gas cooled reactor fuel particles

    International Nuclear Information System (INIS)

    Soo, P.; Uneberg, G.; Sabatini, R.L.; Schweitzer, D.G.

    1979-01-01

    BISO coated UO 2 and ThO 2 particles were heated to high temperatures to determine failure mechanisms during hypothetical loss of coolant scenarios. Rapid failure begins when the oxides are reduced to liquid carbides. Several failure mechanisms are applicable, ranging from hole and crack formation in the coatings to catastrophic particle disintegration

  11. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  12. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  13. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  14. Preliminary analysis of combined cycle of modular high-temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Baogang, Z.; Xiaoyong, Y.; Jie, W.; Gang, Z.; Qian, S.

    2015-01-01

    Modular high-temperature gas cooled reactor (HTGR) is known as one of the most advanced nuclear reactors because of its inherent safety and high efficiency. The power conversion system of HTGR can be steam turbine based on Rankine cycle or gas turbine based on Brayton cycle respectively. The steam turbine system is mature and the gas turbine system has high efficiency but under development. The Brayton-Rankine combined cycle is an effective way to further promote the efficiency. This paper investigated the performance of combined cycle from the viewpoint of thermodynamics. The effect of non-dimensional parameters on combined cycle’s efficiency, such as temperature ratio, compression ratio, efficiency of compressor, efficiency of turbine, was analyzed. Furthermore, the optimal parameters to achieve highest efficiency was also given by this analysis under engineering constraints. The conclusions could be helpful to the design and development of combined cycle of HTGR. (author)

  15. Modular High Temperature Gas-Cooled Reactor heat source for coal conversion

    International Nuclear Information System (INIS)

    Schleicher, R.W. Jr.; Lewis, A.C.

    1992-09-01

    In the industrial nations, transportable fuels in the form of natural gas and petroleum derivatives constitute a primary energy source nearly equivalent to that consumed for generating electric power. Nations with large coal deposits have the option of coal conversion to meet their transportable fuel demands. But these processes themselves consume huge amounts of energy and produce undesirable combustion by-products. Therefore, this represents a major opportunity to apply nuclear energy for both the environmental and energy conservation reasons. Because the most desirable coal conversion processes take place at 800 degree C or higher, only the High Temperature Gas-Cooled Reactors (HTGRs) have the potential to be adapted to coal conversion processes. This report provides a discussion of this utilization of HTGR reactors

  16. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    International Nuclear Information System (INIS)

    Forsberg, Charles; Hu, Lin-wen; Peterson, Per; Sridharan, Kumar

    2015-01-01

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  17. Computational model for a high temperature electrolyzer coupled to a HTTR for efficient nuclear hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Daniel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy; Gamez, Abel; Brayner, Carlos, E-mail: danielgonro@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Garcia, Lazaro; Garcia, Carlos; Torre, Raciel de la, E-mail: lgarcia@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Sanchez, Danny [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

    2015-07-01

    High temperature electrolysis process coupled to a very high temperature reactor (VHTR) is one of the most promising methods for hydrogen production using a nuclear reactor as the primary heat source. However there are not references in the scientific publications of a test facility that allow to evaluate the efficiency of the process and other physical parameters that has to be taken into consideration for its accurate application in the hydrogen economy as a massive production method. For this lack of experimental facilities, mathematical models are one of the most used tools to study this process and theirs flowsheets, in which the electrolyzer is the most important component because of its complexity and importance in the process. A computational fluid dynamic (CFD) model for the evaluation and optimization of the electrolyzer of a high temperature electrolysis hydrogen production process flowsheet was developed using ANSYS FLUENT®. Electrolyzer's operational and design parameters will be optimized in order to obtain the maximum hydrogen production and the higher efficiency in the module. This optimized model of the electrolyzer will be incorporated to a chemical process simulation (CPS) code to study the overall high temperature flowsheet coupled to a high temperature accelerator driven system (ADS) that offers advantages in the transmutation of the spent fuel. (author)

  18. Computational model for a high temperature electrolyzer coupled to a HTTR for efficient nuclear hydrogen production

    International Nuclear Information System (INIS)

    Gonzalez, Daniel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy; Gamez, Abel; Brayner, Carlos; Garcia, Lazaro; Garcia, Carlos; Torre, Raciel de la; Sanchez, Danny

    2015-01-01

    High temperature electrolysis process coupled to a very high temperature reactor (VHTR) is one of the most promising methods for hydrogen production using a nuclear reactor as the primary heat source. However there are not references in the scientific publications of a test facility that allow to evaluate the efficiency of the process and other physical parameters that has to be taken into consideration for its accurate application in the hydrogen economy as a massive production method. For this lack of experimental facilities, mathematical models are one of the most used tools to study this process and theirs flowsheets, in which the electrolyzer is the most important component because of its complexity and importance in the process. A computational fluid dynamic (CFD) model for the evaluation and optimization of the electrolyzer of a high temperature electrolysis hydrogen production process flowsheet was developed using ANSYS FLUENT®. Electrolyzer's operational and design parameters will be optimized in order to obtain the maximum hydrogen production and the higher efficiency in the module. This optimized model of the electrolyzer will be incorporated to a chemical process simulation (CPS) code to study the overall high temperature flowsheet coupled to a high temperature accelerator driven system (ADS) that offers advantages in the transmutation of the spent fuel. (author)

  19. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  20. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  1. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  3. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  5. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  6. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

    2003-01-01

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  7. Selection of design basis event for modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

    2016-06-01

    Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

  8. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  10. The development of high-temperature reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Engelmann, P.; Krings, F.

    1980-01-01

    The principal features of high-temperature reactors are recalled, then the current state of technology of the line in the Federal Republic of Germany is described. Reference is made to the experience of operating the AVR reactor, the construction of the THTR-300 reactor as well as the HTT and PNP projects [fr

  11. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  12. Research activities on high-temperature gas-cooled reactors (HTRs) in the 5. EURATOM RTD Framework programme

    International Nuclear Information System (INIS)

    Martin-Bermejo, J.; Hugon, M.; Van Goethem, G.

    2002-01-01

    One of the areas of research of the 'nuclear fission' key action of the 5. EURATOM RTD Framework Programme (FP5) is the safety and efficiency of future systems. The main objective of this area is to investigate and evaluate new or revisited concepts (both reactors and alternative fuels) for nuclear energy that offer potential longer term benefits in terms of cost, safety, waste management, use of fissile material, less risk of diversion and sustainability. Several projects related to high-temperature gas-cooled reactors (HTRs) were retained by the European Commission (EC) services. They address important issues such as HTR fuel technology, HTR fuel cycle, HTR materials, power conversion systems and licensing. Most of these projects have already started and are progressing according to the schedule. They are the initial core of activities of a European Network on 'High-temperature Reactor Technology' (HTR-TN) recently set up by 18 EU organisations. (authors)

  13. Properties of super alloys for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Izaki, Takashi; Nakai, Yasuo; Shimizu, Shigeki; Murakami, Takashi

    1975-01-01

    The existing data on the properties at high temperature in helium gas of iron base super alloys. Incoloy-800, -802 and -807, nickel base super alloys, Hastelloy-X, Inconel-600, -617 and -625, and a casting alloy HK-40 were collectively evaluated from the viewpoint of the selection of material for HTGRs. These properties include corrosion resistance, strength and toughness, weldability, tube making, formability, radioactivation, etc. Creep strength was specially studied, taking into consideration the data on the creep characteristics in the actual helium gas atmosphere. The necessity of further long run creep data is suggested. Hastelloy-X has completely stable corrosion resistance at high temperature in helium gas. Incoloy 800 and 807 and Inconel 617 are not preferable in view of corrosion resistance. The creep strength of Inconel 617 extraporated to 1,000 deg C for 100,000 hours in air was the greatest rupture strength of 0.6 kg/mm 2 in all above alloys. However, its strength in helium gas began to fall during a relatively short time, so that its creep strength must be re-evaluated in the use for long time. The radioactivation and separation of oxide film in primary construction materials came into question, Inconel 617 and Incoloy 807 showed high induced radioactivity intensity. Generally speaking, in case of nickel base alloys such as Hastelloy-X, oxide film is difficult to break away. (Iwakiri, K.)

  14. Consideration of ultra-high temperature nuclear heat sources for MHD conversion systems

    International Nuclear Information System (INIS)

    Holman, R.R.; Tobin, J.M.; Young, W.E.

    1975-01-01

    The nuclear technology reactors developed and tested in the Nuclear Engine Rocket Vehicle Application (NERVA) program operated with fuel exit gas temperatures in excess of 2600 K. This experience provided a significant ultra-high temperature technology base and design insight for commercial power applications. Design approaches to accommodate fission product retention and other key prevailing requirements are examined in view of the basic overriding functional requirements, and some interesting reconsiderations are suggested. Predicted overall system performance potentials for a 2000 K MHD conversion system and reactor parameter requirements are compared and related to existing technology status. Needed verification and development efforts are suggested. A reconsideration of basic design approaches is suggested that could open the door for immediate development of ultrahigh temperature nuclear heat sources for advanced energy systems

  15. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

  16. Advanced multi-physics simulation capability for very high temperature reactors

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

    2012-01-01

    The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

  17. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  18. Plutonium and minor actinides management in thermal high - temperature reactors - the EU FP6 project puma

    International Nuclear Information System (INIS)

    Kuijper, J. C.

    2007-01-01

    The High Temperature gas-cooled Reactor (HTR) can fulfil a very useful niche for the purposes of Pu and Minor Actinide (MA) incineration due to its unique and unsurpassed safety features, as well as to the attractive incentives offered by the nature of the coated particle (CP) fuel. No European reactor of this type is currently available, but there has been, and still is, considerable interest internationally. Decisions to construct such a reactor in China and in South Africa have already been made or are about to be made. Apart from the unique and unsurpassed safety features offered by this reactor type, the nature of the CP fuel offers a number of attractive characteristics. In particular, it can withstand burn-ups far beyond that in either LWR or FR systems. Demonstrations as high as 75% FIMA have been achieved. The coated particle itself offers significantly improved proliferation resistance, and finally with a correct choice of the kernel composition, it can be a very effective support for direct geological disposal of the fuel. The overall objective of the PUMA project, a Specific Targeted Research Project (STREP) within the European Union 6th Framework (EU FP6), is to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO 2 -free energy generation. A number of important issues concerning the use of Pu and MA in gas-cooled reactors have already been dealt with in other projects, or are being treated in ongoing projects, e.g. as part of EU FP6. However, further steps are required to demonstrate the potential of HTRs as Pu/MA transmuters based on realistic/feasible designs of CP Pu/MA fuel and the PUMA focuses on necessary

  19. The Dragon project and high temperature reactor (HTR position)

    International Nuclear Information System (INIS)

    Shepherd, L.

    1981-01-01

    After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  1. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, Winfried

    1997-01-01

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  2. Utility/user requirements for the modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Boyer, V.S.; Kendall, J.M.; Gotschall, H.L.

    1989-01-01

    This paper describes the approach used by Gas-Cooled Reactor Associates (GCRA) in developing Utility/User Requirements for the Modular High Temperature Gas-cooled Reactor (MHTGR). As representatives of the Utility/User industry, it is GCRA's goal that the MHTGR concept be established as an attractive nuclear option offering competitive economics and limited ownership risks. Commercially deployed MHTGR systems should then compete favorably in a mixed-fuel economy with options using fossil, other nuclear and other non-fossil sources. To achieve this goal, the design of the MHTGR plant must address the problems experienced by the U.S. industrial infrastructure during deployment of the first generation of nuclear plants. Indeed, it is GCRA's intent to utilize the characteristics of MHTGR technology for the development of a nuclear alternative that poses regulatory, financial and operational demands on the Owner/Operator that are, in aggregate, comparable to those encountered with non-nuclear options. The dominant risks faced by U.S. Utilities with current nuclear plants derive from their operational complexity and the degree of regulatory involvement in virtually all aspects of utility operations. The MHTGR approach of using ceramic fuel coatings to contain fission products provides the technical basis for simplification of the plant and stabilization of licensing requirements and thus the opportunity for reducing the risks of nuclear plant ownership. The paper describes the rationale for the selection of key requirements for public safety, plant size and performance, operations and maintenance, investment protection, economics and siting in the context of a risk management philosophy. It also describes the ongoing participation of the Utility/User in interpreting requirements, conducting program and design reviews and establishing priorities from the Owner/Operator perspective. (author). 7 refs, 1 fig

  3. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Marmier, Alain

    2012-01-01

    -state conditions, both particle failure fractions (calculated with the CRYSTAL code) and the fissile material cost were determined: Wallpaper type of fuel impacts positively on the fuel cycle and reduces both the need for fissile material and the production of minor actinides, facilitating fuel reprocessing and reducing fuel cost. Safety is also improved with particle temperature being reduced during steady-state operation. This reduces the expected particle failure fraction by up to 85% over its in-core lifetime, and also the fission product release. In the long term, the VHTR is believed to be the most suitable concept for co-generation of process heat. However, in recent years the tendency in international projects goes back to lower reactor outlet temperatures mainly for 3 reasons: (1) two of the main driver countries of the VHTR have dropped thermochemical hydrogen production from their high priority list, due to expected economic and material corrosion issues; (2) the extremely high reactor outlet temperatures of > 950 degrees C would require totally new materials and construction standards; (3) the existing process heat market (steam < 600 degrees C) is already so big that investment into longer term concepts receives lower priority. In addition, the high temperature operation would tend to increase particle failure fraction and fission product release. In order to reach the VHTR objectives of high temperature and power conversion efficiency much earlier, a novel approach was developed: The nuclear part of the power plant would run at acceptably low temperature and would power a compression heat pump system acting as a temperature booster. Thus, very high temperature operation could be limited to a section of a conventional external gas circuit, avoiding the constraints related to the combination of very high temperatures and irradiation. (author)

  4. Alcohol synthesis in a high-temperature slurry reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, G.W.; Marquez, M.A.; McCutchen, M.S. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system can be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  6. Evaluation of Candidate Linear Variable Displacement Transducers for High Temperature Irradiations in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Knudson, D.L.; Rempe, J.L.; Daw, J.E.

    2009-01-01

    The United States (U.S.) Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to promote nuclear science and technology in the U.S. Given this designation, the ATR is supporting new users from universities, laboratories, and industry as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A fundamental component of the ATR NSUF program is to develop in-pile instrumentation capable of providing real-time measurements of key parameters during irradiation experiments. Dimensional change is a key parameter that must be monitored during irradiation of new materials being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can experience significant changes during high temperature irradiation. Currently, dimensional changes are determined by repeatedly irradiating a specimen for a defined period of time in the ATR and then removing it from the reactor for evaluation. The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data (i.e., only characterizing the end state when samples are removed from the reactor) and may disturb the phenomena of interest. To address these issues, the Idaho National Laboratory (INL) recently initiated efforts to evaluate candidate linear variable displacement transducers (LVDTs) for use during high temperature irradiation experiments in typical ATR test locations. Two nuclear grade LVDT vendor designs were identified for consideration - a smaller diameter design qualified for temperatures up to 350 C and a larger design with capabilities to 500 C. Initial evaluation efforts include collecting calibration data as a function of temperature, long duration testing of LVDT response while held at high temperature, and the assessment of changes

  7. Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, W.

    1996-01-01

    This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

  8. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  9. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  10. The real gas behaviour of helium as a cooling medium for high-temperature reactors

    International Nuclear Information System (INIS)

    Hewing, G.

    1977-01-01

    The article describes the influence of the real gas behaviour on the variables of state for the helium gas and the effects on the design of high-temperature reactor plants. After explaining the basic equations for describing variables and changes of state of the real gas, the real and ideal gas behaviour is analysed. Finally, the influence of the real gas behaviour on the design of high-temperature reactors in one- and two-cycle plants is investigated. (orig.) [de

  11. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

  12. The gas-cooled high temperature reactor: perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

  13. High temperature gas cooled reactor technology development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-12-01

    The successful introduction of an advanced nuclear power plant programme depends on many key elements. It must be economically competitive with alternative sources of energy, its technical development must assure operational dependability, the support of society requires that it be safe and environmentally acceptable, and it must meet the regulatory standards developed for its use and application. These factors interrelate with each other, and the ability to satisfy the established goals and criteria of all of these requirements is mandatory if a country or a specific industry is to proceed with a new, advanced nuclear power system. It was with the focus on commercializing the high temperature gas cooled reactor (HTGR) that the IAEA's International Working Group on Gas Cooled Reactors recommended this Technical Committee Meeting (TCM) on HTGR Technology Development. Over the past few years, many Member States have instituted a re-examination of their nuclear power policies and programmes. It has become evident that the only realistic way to introduce an advanced nuclear power programme in today's world is through international co-operation between countries. The sharing of expertise and technical facilities for the common development of the HTGR is the goal of the Member States comprising the IAEA's International Working Group on Gas Cooled Reactors. This meeting brought together key representatives and experts on the HTGR from the national organizations and industries of ten countries and the European Commission. The state electric utility of South Africa, Eskom, hosted this TCM in Johannesburg, from 13 to 15 November 1996. This TCM provided the opportunity to review the status of HTGR design and development activities, and especially to identify international co-operation which could be utilized to bring about the commercialization of the HTGR

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  15. Electrical measuring device for a high temperature reactor

    International Nuclear Information System (INIS)

    Elter, C.; Handel, H.; Schoening, J.; Schmitt, H.

    1982-01-01

    The device for measuring the low or high neutron flux during start-up or at load is accommodated in an armoured guide tube projecting into the floor. A gas-tight capsule is formed as the measuring column with outer dome with a lid solidly connected by a flange to the armoured tube situated on the side wall of the concrete reactor vessel, together with the armoured guide tube. Two shielding shutters prevent the passage of radiation through the armoured tube. (DG) [de

  16. High temperature metallic materials for gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    The Specialists' Meeting was organized in conjunction with an earlier meeting on this topic held in Vienna, Austria, 1981, which provided for a comprehensive review of the status of materials development and testing at that time and for a description of test facilities. This meeting provided an opportunity (1) to review and discuss the progress made since 1981 in the development, testing and qualification of high temperature metallic materials, (2) to critically assess results achieved, and (3) to give directions for future research and development programmes. In particular, the meeting provided a form for a close interaction between component designers and materials specialists. The meeting was attended by 48 participants from France, People's Republic of China, Federal Republic of Germany, Japan, Poland, Switzerland, United Kingdom, USSR and USA presenting 22 papers. The technical part of the meeting was subdivided into four technical sessions: Components Design and Testing - Implications for Materials (4 papers); Microstructure and Environmental Compatibility (4 papers); Mechanical Properties (9 papers); New Alloys and Developments (6 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. This volume contains all papers presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  17. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  18. Nuclear data generation for cryogenic moderators and high temperature moderators

    International Nuclear Information System (INIS)

    Petriw, Sergio

    2007-01-01

    The commonly used processing codes for nuclear data only allow the generation of cross section data for a limited number of materials and physical conditions.At present, one of the most used computer codes for the generation of neutron cross sections is N J O Y, which is based on a phonon expansion of the scattering function starting from the frequency spectrum.Therefore, the information related to the system's density of states is crucial to produce the required data of interest. In this work the formalism of the Synthetic Model for Molecular Solids (S M M S) was implemented, which is in turn based on the Synthetic Frequency Spectrum (S F S) concept.The synthetic spectrum is central in the present work, and it is built from simple, relevant parameters of the moderator, thus conforming an alternative tool when no information on the actual frequency spectrum of the moderator material is available.S F S 's for several material of interest where produced in this work, for both cryogenic and high temperature moderators.We studied some materials of special interest, like solid methane, ice, methyl clathrate and two which are of special interest in the nuclear industry: graphite and beryllium.The libraries generated in the present work for the materials considered, in spite of their synthetic origin, are able to produce results that are even in better agreement with available information [es

  19. Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

    International Nuclear Information System (INIS)

    Quadakkers, W.J.

    1985-01-01

    The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

  20. Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

  1. LARGE-SCALE HYDROGEN PRODUCTION FROM NUCLEAR ENERGY USING HIGH TEMPERATURE ELECTROLYSIS

    International Nuclear Information System (INIS)

    O'Brien, James E.

    2010-01-01

    Hydrogen can be produced from water splitting with relatively high efficiency using high-temperature electrolysis. This technology makes use of solid-oxide cells, running in the electrolysis mode to produce hydrogen from steam, while consuming electricity and high-temperature process heat. When coupled to an advanced high temperature nuclear reactor, the overall thermal-to-hydrogen efficiency for high-temperature electrolysis can be as high as 50%, which is about double the overall efficiency of conventional low-temperature electrolysis. Current large-scale hydrogen production is based almost exclusively on steam reforming of methane, a method that consumes a precious fossil fuel while emitting carbon dioxide to the atmosphere. Demand for hydrogen is increasing rapidly for refining of increasingly low-grade petroleum resources, such as the Athabasca oil sands and for ammonia-based fertilizer production. Large quantities of hydrogen are also required for carbon-efficient conversion of biomass to liquid fuels. With supplemental nuclear hydrogen, almost all of the carbon in the biomass can be converted to liquid fuels in a nearly carbon-neutral fashion. Ultimately, hydrogen may be employed as a direct transportation fuel in a 'hydrogen economy.' The large quantity of hydrogen that would be required for this concept should be produced without consuming fossil fuels or emitting greenhouse gases. An overview of the high-temperature electrolysis technology will be presented, including basic theory, modeling, and experimental activities. Modeling activities include both computational fluid dynamics and large-scale systems analysis. We have also demonstrated high-temperature electrolysis in our laboratory at the 15 kW scale, achieving a hydrogen production rate in excess of 5500 L/hr.

  2. Digital Information Platform Design of Fuel Element Engineering For High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Du Yuwei

    2014-01-01

    This product line provide fuel element for high temperature gas-cooled reactor nuclear power plant which is being constructed in Shidao bay in Shandong province. Its annual productive capacity is thirty ten thousands fuel elements whose shape is spherical . Compared with pressurized water fuel , this line has the feature of high radiation .In order to reduce harm to operators, the comprehensive information platform is designed , which can realize integration of automation and management for plant. This platform include two nets, automation net using field bus technique and information net using Ethernet technique ,which realize collection ,control, storage and publish of information.By means of construction, automatization and informatization of product line can reach high level. (author)

  3. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Jianghai Li

    2017-01-01

    Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

  4. MOTHER MK II: An advanced direct cycle high temperature gas reactor

    International Nuclear Information System (INIS)

    Hart, R.S.; Kendall, J.M.; Marsden, B.J.

    2003-01-01

    The MOTHER (MOdular Thermal HElium Reactor) power plant concepts employ high temperature gas reactors utilizing TRISO fuel, graphite moderator, and helium coolant, in combination with a direct Brayton cycle for electricity generation. The helium coolant from the reactor vessel passes through a Power Conversion Unit (PCU), which includes a turbine-generator, recuperator, precooler, intercooler and turbine-compressors, before being returned to the reactor vessel. The PCU substitutes for the reactor coolant system pumps and steam generators and most of the Balance Of Plant (BOP), including the steam turbines and condensers, employed by conventional nuclear power plants utilizing water cooled reactors. This provides a compact, efficient, and relatively simple plant configuration. The MOTHER MK I conceptual design, completed in the 1987 - 1989 time frame, was developed to economically meet the energy demands for extracting and processing heavy oil from the tar sands of western Canada. However, considerable effort was made to maximize the market potential beyond this application. Consistent with the remote and very high labour rate environment in the tar sands region, simplification of maintenance procedures and facilitation of 'change-out' in lieu of in situ repair was a design focus. MOTHER MK I had a thermal output of 288 MW and produced 120 MW electrical when operated in the electricity only production mode. An annular Prismatic reactor core was utilized, largely to minimize day-to-day operations activities. Key features of the power conversion system included two Power Conversion Units (144 MW th each), the horizontal orientation of all rotating machinery and major heat exchangers axes, high speed rotating machinery (17,030 rpm for the turbine-compressors and 10,200 rpm for the power turbine-generator), gas (helium) bearings for all rotating machinery, and solid state frequency conversion from 170 cps (at full power) to the grid frequency. Recognizing that the on

  5. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

  6. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  7. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  8. The further advancement of the high temperature reactor line

    International Nuclear Information System (INIS)

    Oehme, H.; Schloesser, J.

    1978-01-01

    The development of the HTR line has been reoriented since early 1976 in the light of five objectives: the more than ten concept variants of electricity and process heat generating HTR's were reduced to one concept each. Both concepts agree in their main basic elements, using spherical fuel elements, prestressed concrete pressure vessels and integrated designs. For electricity generation the HTR with a direct coupled gas turbine (HHT system) was chosen in order to achieve an optimum combination of forces of operators, manufacturers and in research the project structure was streamlined. Project companies of the operators are presently being established; the GHT and HBR companies will set up a consortium for planning the nuclear heat supply system. Time scheduling was concentrated on the first HHT and process heat plants. A tender for the HHT power plant is to be submitted in 1983, for the PNP plant in 1985. The THTR-300 nuclear power plant is due to be completed in 1981. The technical status of the HTR projects is characterized as follows: satisfactory progress has been achieved in the THTR-300 in 1977. Three important permits under the Atomic Energy Act are due to be issued in 1978 and, in addition, an application is to be filed for the first partial construction permit. In 1977 the AVR Experimental Nuclear Power Station had been in operation for ten years and continues to be operated successfully at 950 0 C. In the Nuclear Process Heat Prototype Plant Project (PNP) a number of components were harmonized with those of the electricity producing HTR in 1977. The basic concept of the prototype facility was revamped. All basic development work is continued as planned. In the gasification sector a successful demonstration has been launched of the hydrogenating gasification of lignite. (orig.) [de

  9. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  10. The modular high-temperature gas-cooled reactor - a new production reactor

    International Nuclear Information System (INIS)

    Nulton, J.D.

    1990-01-01

    One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

  11. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

  12. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  13. Synthetic fuel production using Texas lignite and a very high temperature reactor for process heat

    International Nuclear Information System (INIS)

    Ross, M.A.; Klein, D.E.

    1982-01-01

    Two approaches for synthetic fuel production from coal are studied using Texas lignite as the feedstock. First, the gasification and liquefaction of coal are accomplished using Lurgi gasifiers and Fischer-Tropsch synthesis. A 50 000 barrel/day facility, consuming 13.7 million tonne/yr (15.1 million ton/yr) of lignite, is considered. Second, a nuclear-assisted coal conversion approach is studied using a very high temperature gas-cooled reactor with a modified Lurgi gasifier and Fischer-Tropsch synthesis. The nuclear-assisted approach resulted in a 35% reduction in coal consumption. In addition, process steam consumption was reduced by one-half and the oxygen plants were eliminated in the nuclear assisted process. Both approaches resulted in a synthetic oil price higher than the March 1980 imported price of $29.65 per barrel: $36.15 for the lignite-only process and $35.16 for the nuclear-assisted process. No tax advantage was assumed for either process and the utility financing method was used for both economic calculations

  14. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  15. The design status of the United States Department of Energy modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Mills, Raymond R. Jr.

    1990-01-01

    The U.S. Department of Energy's Modular High Temperature Gas Cooled Reactor (MHTGR) is being designed using a systems engineering approach referred to as the integrated approach. The top level requirement for the plant is that it provides safe, reliable, economical energy. The safety requirements are established by the U.S. Licensing Authorities, principally the Nuclear Regulatory Commission. The reliability and economic requirements associated with the top level functions have been established in close coordination and cooperation with the electrical utilities and other potential users, and the nuclear supply industry. The integrated approach uses functional analysis to define the functions and sub-functions for the plant and to identify quantitatively how the various functions must be fulfilled. The top four functions associated with the MHTGR are: maintain safe plant operation; maintain plant protection; maintain control of radionuclide release; maintain emergency preparedness. In addition to meeting all U.S. Regulatory Requirements this advanced reactor concept is being designed to meet the following requirements: do not require sheltering or evacuating of anyone outside the plant boundary of 425 meters as a result of normal or abnormal plant operation; do not require operator action in order to accomplish the above sheltering and evacuation objectives and the design must be insensitive to operator errors; utilize inherent characteristics of materials to develop passive safety features; provide very long times for corrective actions following the initiation of an abnormal event before plant damage would be incurred

  16. Criticality assessment for prismatic high temperature reactors by fuel stochastic Monte Carlo modeling

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    Modeling of prismatic high temperature reactors requires a high precision description due to the triple heterogeneity of the core and also to the random distribution of fuel particles inside the fuel pins. On the latter issue, even with the most advanced Monte Carlo techniques, some approximation often arises while assessing the criticality level: first, a regular lattice of TRISO particles inside the fuel pins and, second, the cutting of TRISO particles by the fuel boundaries. We utilized two of the most accurate Monte Codes: MONK and MCNP, which are both used for licensing nuclear power plants in United Kingdom and in the USA, respectively, to evaluate the influence of the two previous approximations on estimating the criticality level of the Gas Turbine Modular Helium Reactor. The two codes exactly shared the same geometry and nuclear data library, ENDF/B, and only modeled different lattices of TRISO particles inside the fuel pins. More precisely, we investigated the difference between a regular lattice that cuts TRISO particles and a random lattice that axially repeats a region containing over 3000 non-cut particles. We have found that both Monte Carlo codes provide similar excesses of reactivity, provided that they share the same approximations.

  17. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Vasudevan, Vijay; Carroll, Laura; Sham, Sam

    2015-01-01

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  18. Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

    Science.gov (United States)

    Peterman, D. D.; Fontaine, R. W.; Quade, R. N.; Halvers, L. J.; Jahromi, A. M.

    1975-01-01

    The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

  19. Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

    International Nuclear Information System (INIS)

    Peterman, D.D.; Fontaine, R.W.; Quade, R.N.; Halvers, L.J.; Jahromi, A.M.

    1975-01-01

    The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program

  20. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, Vijay [Univ. of Cincinnati, OH (United States); Carroll, Laura [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-06

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  2. Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Panda, Binayak; Shah, Sandeep

    2005-01-01

    Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

  3. High temperature brazing of primary-system components in the nuclear field

    International Nuclear Information System (INIS)

    Belicic, M.; Fricker, H.W.; Iversen, K.; Leukert, W.

    1981-01-01

    Apart from the well-known welding procedures, high-temperature brazing is successfully applied in the manufacture of primary components in the field of nuclear reactor construction. This technique is applied in all cases where apart from sufficient resistance and high production safety importance is laid on dimensional stability without subsequent mechanical processing of the components. High-temperature brazing is therefore very important in the manufacture of fuel rod spacers or control rod guide tubes. In this context, during one brazing process many brazing seams have to be produced in extremely narrow areas and within small tolerances. As basic materials precipitation hardening alloys with a high nickel percentage, austenitic Cr-Ni-steels or the zirconium alloy Zry 4 are used. Generally applied are: boron free nickel or zirconium brazing filler metals. (orig.)

  4. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    International Nuclear Information System (INIS)

    Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

    2007-01-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold (1) efficient low cost energy generation and (2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  5. Hydrogen production system based on high temperature gas cooled reactor energy using the sulfur-iodine (SI) thermochemical water splitting cycle

    International Nuclear Information System (INIS)

    Garcia, L.; Gonzalez, D.

    2011-01-01

    Hydrogen production from water using nuclear energy offers one of the most attractive zero-emission energy strategies and the only one that is practical on a substantial scale. Recently, strong interest is seen in hydrogen production using heat of a high-temperature gas-cooled reactor. The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using thermochemical or high-temperature electrolysis (HTE) processes. Eventually it could be also employ a high-temperature gas-cooled reactor (HTGR), which is particularly attractive because it has unique capability, among potential future generation nuclear power options, to produce high-temperature heat ideally suited for nuclear-heated hydrogen production. Using heat from nuclear reactors to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been interest of many laboratories in the world. One of the promising approaches to produce large quantity of hydrogen in an efficient way using the nuclear energy is the sulfur-iodine (SI) thermochemical water splitting cycle. Among the thermochemical cycles, the sulfur iodine process remains a very promising solution in matter of efficiency and cost. This work provides a pre-conceptual design description of a SI-Based H2-Nuclear Reactor plant. Software based on chemical process simulation (CPS) was used to simulate the thermochemical water splitting cycle Sulfur-Iodine for hydrogen production. (Author)

  6. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  7. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-01-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  8. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  9. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  10. Coated particle fuel for high temperature gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

    2007-10-15

    applications at 850-900 .deg. C and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 {mu}m diameter UO{sub 2} kernel of 10% enrichment is surrounded by a 100 {mu}m thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 {mu}m thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level.

  11. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

    2007-01-01

    and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO 2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  12. Present state and future prospect of development of high temperature gas-cooled reactors in Japan

    International Nuclear Information System (INIS)

    Sanokawa, Konomo

    1994-01-01

    High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

  13. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  14. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  15. Determination of an instability temperature for alloys in the cooling gas of a high temperature reactor

    International Nuclear Information System (INIS)

    Grimmer, H.; Grman, D.; Krompholz, K.; Zimmermann, U.; Ullrich, G.

    1985-05-01

    High temperature alloys designed to be used for components in the primary circuit of a helium cooled high temperature nuclear reactor show massive CO production above a certain temperature, called the instability temperature T/sub i/, which increases with increasing partial pressure of CO in the cooling gas. At p/sub CO/ = 15 microbar, T/sub i/ lies between 900 and 950 degrees C for the four alloys under investigation: T/sub i/ is lowest for the iron base alloy Incoloy 800 H and increases for the nickel base alloys in the order Inconel 617, HDA 230 and Nimonic 86. Measurements of T/sub i/ made at 3 different laboratories were compared and shown to agree for p/sub CO/ 25 microbar, compatible with CO production by a reaction of Cr2O3 with carbides. Some measurements of T/sub i/ on HDA 230 and Nimonic 86 were performed in the course of simulated reactor disturbances. They showed that the oxide layer looses its protective properties above T/sub i/. A highlight of the examinations was the detection of eta-carbides (M6C) with unusual properties. M6C is the only type of carbide occuring in HDA 230. An eta-carbide with a lattice constant of 1088.8 pm had developed at the surface of Nimonic 86 during pre-oxidation before the disturbance simulation. Its composition is estimated at Ni3SiMo2C. Eta-carbides containing Si and especially eta-carbides with lattice constants as low as 1088.8 pm have been described only rarely until now. (author)

  16. Feasibility study on silicon doping using high temperature test engineering reactor

    International Nuclear Information System (INIS)

    Seki, Masaya; Takaki, Naoyuki; Goto, Minoru; Shimakawa, Satoshi

    2011-01-01

    The feasibility study on silicon doping using the High Temperature engineering Test Reactor (HTTR) is performed by numerical simulations. The HTTR is a High Temperature Gas-cooled Reactor (HTGR) situated at JAEA Oarai research and development center. It has a 30MW thermal power and the outlet coolant temperature is 950degC. The objective of this study is to evaluate the following issues, 1. The impact of loading Si-ingots into the core on the criticality, 2. The uniformity of the neutron capture reaction rate in Si-ingots, and 3. The production rate of silicon semiconductor. In this study, six Si-ingots are loaded into the irradiation area which is located in the peripheral region of the core. They are irradiated with rotation movement around the axial direction to obtain uniform neutron capture reaction rate in the radial direction. Additionally, the neutron filter, which is made of graphite containing boron, is used to obtain uniform neutron capture reaction rate in the axial direction. The evaluations were conducted by performing the HTTR whole core calculations with the Monte Carlo code MVP-2.0. In the calculations, several tally regions were defined on the Si-ingots to investigate the uniformity of the neutron capture reaction rate. As a result, loading the Si-ingots into the core causes negative reactivity by about 0.7%dk/k. Uniform neutron capture reaction rate of Si-ingot is obtained 98% in the radial and the axial direction. In case of the target of semiconductor resistivity is set to 50 Ωcm, the required irradiation time becomes 10 hours. The HTTR is able to produce silicon semiconductor of 540kg in one-time irradiation. This study was conducted as a joint research with JAEA, Nuclear Fuel Industries, LTD, Toyota Tsusho Corporation and Tokai University. (author)

  17. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR

    International Nuclear Information System (INIS)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L.

    2008-01-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise b enchmark . The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or p itch o f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  18. The gas-cooled high temperature reactor. Perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years extensive research and development programmes on helium-cooled high temperature reactors (HTR) have been carried out in several countries of the world. As a result of the long-standing efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high temperature fuels and materials technology. Three small experimental plants have been operated successfully over extended periods of time. Prototype steam-cycle plants of 300MW(e) are under way at Fort St. Vrain (full-power operation scheduled for 1977) and at Schmehausen (scheduled for 1979). Major delays have occurred in the construction and commissioning of these plants for various reasons but do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successful. Major efforts both by governments and industry are now required to ensure a successful second approach. To reach competitivity with established nuclear power systems and to take full advantage of the fuel conservation potential of the HTR requires the implementation of the closed thorium fuel cycle on a commercial scale. While some key steps of this cycle have been implemented on a laboratory scale, progress towards a prototype recycling facility has been slow. Closing the thorium fuel cycle represents a major challenge and can only be achieved in a close international collaboration. The paper discusses the world-wide status and potential of HTR technology and reviews the major international development programmes. (author)

  19. High-temperature reactors for underground liquid-fuels production with direct carbon sequestration

    International Nuclear Information System (INIS)

    Forsberg, C. W.

    2008-01-01

    The world faces two major challenges: (1) reducing dependence on oil from unstable parts of the world and (2) minimizing greenhouse gas emissions. Oil provides 39% of the energy needs of the United States, and oil refineries consume over 7% of the total energy. The world is running out of light crude oil and is increasingly using heavier fossil feedstocks such as heavy oils, tar sands, oil shale, and coal for the production of liquid fuels (gasoline, diesel, and jet fuel). With heavier feedstocks, more energy is needed to convert the feedstocks into liquid fuels. In the extreme case of coal liquefaction, the energy consumed in the liquefaction process is almost twice the energy value of the liquid fuel. This trend implies large increases in carbon dioxide releases per liter of liquid transport fuel that is produced. It is proposed that high-temperature nuclear heat be used to refine hydrocarbon feedstocks (heavy oil, tar sands, oil shale, and coal) 'in situ ', i.e., underground. Using these resources for liquid fuel production would potentially enable the United States to become an exporter of oil while sequestering carbon from the refining process underground as carbon. This option has become potentially viable because of three technical developments: precision drilling, underground isolation of geological formations with freeze walls, and the understanding that the slow heating of heavy hydrocarbons (versus fast heating) increases the yield of light oils while producing a high-carbon solid residue. Required peak reactor temperatures are near 700 deg. C-temperatures within the current capabilities of high-temperature reactors. (authors)

  20. Numerical evaluation of flow through a prismatic very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Barros Filho, Jose A.; Santos, Andre A.C.; Navarro, Moyses A.; Ribeiro, Felipe Lopes

    2011-01-01

    The High-temperature Gas-cooled reactor (HTGR) is a Next Generation Nuclear System that has a good chance to be used as energy generation source in the near future owing to its potential capacity to supply hydrogen without greenhouse gas emission for the future humanity. Recently, improvements in the HTGR design led to the Very High Temperature Reactor (VHTR) concept in which the outlet temperature of the coolant gas reaches to 1000 deg C increasing the efficiency of the hydrogen and electricity generation. Among the core concepts emerging in the VHTR development stands out the prismatic block which uses coated fuel microspheres named TRISO pressed into cylinders and assembled in hexagonal graphite blocks staked to form columns. The graphite blocks contain flow channels around the fuel cylinders for the helium coolant. In this study an analysis is performed using the CFD code CFX 13.0 on a prismatic fuel assembly in order to investigate its thermo-fluid dynamic performance. The simulations were made in a 1/12 fuel element model of the GT-MHR design which was developed by General Atomics. A numerical mesh verification process based on the Grid Convergence Index (GCI) was performed using five progressively refined meshes to assess the numerical uncertainty of the simulation and determine adequate mesh parameters. An analysis was also performed to evaluate different methods to define the inlet and outlet boundary conditions. In this study simulations of models with and without inlet and outlet plena were compared, showing that the presence of the plena offers a more realistic flow distribution. (author)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  3. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  4. The investigation for attaining the optimal yield of oil shale by integrating high temperature reactors

    International Nuclear Information System (INIS)

    Bhattacharyya, A.T.

    1984-03-01

    This work presents a systemanalytical investigation and shows how far a high temperature reactor can be integrated for achieving the optimal yield of kerogen from oil shale. About 1/3 of the produced components must be burnt out in order to have the required high temperature process heat. The works of IGT show that the hydrogen gasification of oil shale enables not only to reach oil shale of higher quality but also allows to achieve a higher extraction quantity. For this reason a hydro-gasification process has been calculated in this work in which not only hydrogen is used as the gasification medium but also two high temperature reactors are integrated as the source of high temperature heat. (orig.) [de

  5. Rules for design of Alloy 617 nuclear components to very high temperatures

    International Nuclear Information System (INIS)

    Corum, J.M.; Blass, J.J.

    1991-01-01

    Very-high-temperature gas-cooled reactors provide attractive options for electric power generation using a direct gas-turbine cycle and for process-heat applications. For the latter, temperatures of at least 950 degree C (1742 degree F) are desirable. As a first step to providing rules for the design of nuclear components operating at very high temperatures, a draft ASME Boiler and Pressure Vessel Code Case has been prepared by an ad hoc Code task force. The Case, which is patterned after the high-temperature nuclear Code Case N-47, covers Ni-Cr-Co-Mo Alloy 617 for temperatures to 982 degree C (1800 degree F). The purpose of this paper is to provide a synopsis of the draft Case and the significant differences between it and Case N-47. Particular emphasis is placed on the material behavior and allowables. The paper also recommends some materials and structures development activities that are needed to place the design methodology on a sound and defensible footing. 4 refs., 9 figs., 1 tab

  6. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  7. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  8. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  9. High Temperature Reactors for a new IAEA Coordinated Research Project on energy neutral mineral development processes

    Energy Technology Data Exchange (ETDEWEB)

    Haneklaus, Nils, E-mail: n.haneklaus@berkeley.edu [Department of Nuclear Engineering, University of California, Berkeley, 4118 Etcheverry Hall, MC 1730, Berkeley, CA 94720-1730 (United States); Reitsma, Frederik [IAEA, Division of Nuclear Power, Section of Nuclear Power Technology Development, VIC, PO Box 100, Vienna 1400 (Austria); Tulsidas, Harikrishnan [IAEA, Division of Nuclear Fuel Cycle and Waste Technology, Section of Nuclear Fuel Cycle and Materials, VIC, PO Box 100, Vienna 1400 (Austria)

    2016-09-15

    The International Atomic Energy Agency (IAEA) is promoting a new Coordinated Research Project (CRP) to elaborate on the applicability and potential of using High Temperature Reactors (HTRs) to provide process heat and/or electricity to power energy intensive mineral development processes. The CRP aims to provide a platform for cooperation between HTR-developers and mineral development processing experts. Energy intensive mineral development processes with (e.g. phosphate-, gold-, copper-, rare earth ores) or without (e.g. titanium-, aluminum ore) the possibility to recover accompanying uranium and/or thorium that could be developed and used as raw material for nuclear reactor fuel enabling “energy neutral” processing of the primary ore if the recovered uranium and/or thorium is sufficient to operate the greenhouse gas lean energy source used shall be discussed according to the participants needs. This paper specifically focuses on the aspects to be addressed by HTR-designers and developers. First requirements that should be fulfilled by the HTR-designs are highlighted together with the desired outcomes of the research project.

  10. Study on the profitableness of electricity generation with high temperature reactors

    International Nuclear Information System (INIS)

    Kolb, G.

    1978-08-01

    The programme group 'Systemforschung und Technologische Entwicklung' (STE) of the Nuclear Research Centre Juelich in cooperation with the internal institutions 'Projekttraegerschaft Entwicklung von Hochtemperaturreaktoren' (PTH) and 'Institut fuer Reaktorentwicklung' (IRE) on the one hand, and the external partner 'Hochtemperatur-Reactorbau GmbH' (HRB) and 'Gesellschaft fuer Hochtemperaturreactor Technik' (GHT) on the other hand have set up a study on fuel cycle costs, electricity production cost and the economical use as well as uranium resource protection by introduction of high temperature reactors (HTR) with pebble bed core to generate electricity. The pressurized-water reactor (PWR) today on the market serves as comparison. The working results obtained sofar are compiled in the present report. It was particularly noted that - the HTR can economically fully compete with the PWR for electricity generation - the necessary supply of natural uranium for the HTR in open circuit is about one third lower and in the closed circuit, almost two thirds lower than in the corresponding PWR. A further reduction is possible on a long-term basis by highly converting HTW systems. (orig.) [de

  11. Evaluation of Indirect Combined Cycle in Very High Temperature Gas--Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh; Robert Barner; Cliff Davis; Steven Sherman; Paul Pickard

    2006-01-01

    The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling

  12. Working Towards Unified Safety Design Criteria for Modular High Temperature Gas-cooled Reactor Designs

    International Nuclear Information System (INIS)

    Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

    2014-01-01

    The Nuclear Power Development Section of the IAEA recently received approval for a Coordinated Research Project (CRP) to investigate and make proposals on modular High Temperature Gas-cooled Reactor (HTGR) Safety design criteria. It is expected that these criteria would consider past experience and existing safety standards in the light of modular HTGR material and design characteristics to propose safety design criteria. It will consider the deterministic and risk-informed safety design standards that apply to the wide spectrum of Off- normal events under development worldwide for existing and planned HTGRs. The CRP would also take into account lessons from the Fukushima Daiichi accident, clarifying the safety approach and safety evaluation criteria for design and beyond design basis events, including those events that can affect multiple reactor modules and/or are dependent on the application proximate to the plant site. (e. g., industrial process steam/heat). The logical flow of criteria is from the fundamental inherent safety characteristics of modular HTGRs and associated expected performance characteristics, to the safety functions required to ensure those characteristics during the wide spectrum of Off-normal events, and finally to specific criteria related to those functions. This is detailed in the paper with specific examples included of how it may be applied. The results of the CRP will be made available to the member states and HTGR community. (author)

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  14. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  17. Investigations of anticipated transients without scram (ATWS) for the high temperature reactor

    International Nuclear Information System (INIS)

    Heckhoff, H.D.

    1981-10-01

    In this study anticipated transients without scram (ATWS) are investigated for the high temperature reactor, especially for the thorium high temperature reactor (THTR) 300 MWe as an example. It is shown that the two ATWS 'feedwater flow reduction from full power' and 'positive reactivity insertion of 1 mNile/s from 40 per cent power' are the most important transients for the THTR. The additional load caused by the ATWS can be reduced sufficiently by some small modifications of the afterheat removal system. Supplementary precautions are not necessary. In the last part of this study some possibilities to improve the behaviour of the power plant are shown with regard to high temperature reactors of the future, the partial scram as well as some modifications of heating and cooling of the steam generator. (orig.) [de

  18. Hydrogen Production System with High Temperature Electrolysis for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kentaro, Matsunaga; Eiji, Hoashi; Seiji, Fujiwara; Masato, Yoshino; Taka, Ogawa; Shigeo, Kasai

    2006-01-01

    Steam electrolysis with solid oxide cells is one of the most promising methods for hydrogen production, which has the potential to be high efficiency. Its most parts consist of environmentally sound and common materials. Recent development of ceramics with high ionic conductivity suggests the possibility of widening the range of operating temperature with maintaining the high efficiency. Toshiba is constructing a hydrogen production system with solid oxide electrolysis cells for nuclear power plants. Tubular-type cells using YSZ (Yttria-Stabilized- Zirconia) as electrolyte showed good performance of steam electrolysis at 800 to 900 deg C. Larger electrolysis cells with present configuration are to be combined with High Temperature Reactors. The hydrogen production efficiency on the present designed system is expected around 50% at 800 to 900 deg C of operating temperature. For the Fast Reactors, 'advanced cell' with higher efficiency at lower temperature are to be introduced. (authors)

  19. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  20. Transmutation of nuclear waste in nuclear reactors

    International Nuclear Information System (INIS)

    Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

    1996-03-01

    The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

  1. High temperature phase transitions in nuclear fuels of the fourth generation

    International Nuclear Information System (INIS)

    De Bruycker, F.

    2010-01-01

    Understanding the behaviour of nuclear materials in extreme conditions is of prime importance for the analysis of the operation limits of nuclear fuels, and prediction of possible nuclear reactor accidents, relevant to the general objectives of nuclear safety research. The main purpose of this thesis is the study of high temperature phase transitions in nuclear materials, with special attention to the candidate fuel materials for the reactors of the 4. Generation. In this framework, material properties need to be investigated at temperatures higher than 2500 K, where equilibrium conditions are difficult to obtain. Laser heating combined with fast pyrometer is the method used at the European Institute for Transuranium Elements (JRC - ITU). It is associated to a novel process used to determine phase transitions, based on the detection, via a suited low-power (mW) probe laser, of changes in surface reflectivity that may accompany solid/liquid phase transitions. Fast thermal cycles, from a few ms up to the second, under almost container-free conditions and control atmosphere narrow the problem of vaporisation and sample interactions usually meet with traditional method. This new experimental approach has led to very interesting results. It confirmed earlier research for material systems known to be stable at high temperature (such as U-C) and allowed a refinement of the corresponding phase diagrams. But it was also feasible to apply this method to materials highly reactive, thus original results are presented on PuO 2 , NpO 2 , UO 2 -PuO 2 and Pu-C systems. (author)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  3. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  4. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  5. Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

    1997-01-01

    The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

  6. Application of the High Temperature Gas Cooled Reactor to oil shale recovery

    International Nuclear Information System (INIS)

    Wadekamper, D.C.; Arcilla, N.T.; Impellezzeri, J.R.; Taylor, I.N.

    1983-01-01

    Current oil shale recovery processes combust some portion of the products to provide energy for the recovery process. In an attempt to maximize the petroleum products produced during recovery, the potentials for substituting nuclear process heat for energy generated by combustion of petroleum were evaluated. Twelve oil shale recovery processes were reviewed and their potentials for application of nuclear process heat assessed. The High Temperature Gas Cooled Reactor-Reformer/Thermochemical Pipeline (HTGR-R/TCP) was selected for interfacing process heat technology with selected oil shale recovery processes. Utilization of these coupling concepts increases the shale oil product output of a conventional recovery facility from 6 to 30 percent with the same raw shale feed rate. An additional benefit of the HTGR-R/TCP system was up to an 80 percent decrease in emission levels. A detailed coupling design for a typical counter gravity feed indirect heated retorting and upgrading process were described. Economic comparisons prepared by Bechtel Group Incorporated for both the conventional and HTGR-R/TCP recovery facility were summarized

  7. Economic evaluation of the steam-cycle high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1983-07-01

    The High Temperature Gas-Cooled Reactor is unique among current nuclear technologies in its ability to generate energy in temperature regimes previously limited to fossil fuels. As a result, it can offer commercial benefits in the production of electricity, and at the same time, expand the role of nuclear energy to the production of process heat. This report provides an evaluation of the HTGR-Steam Cycle (SC) system for the production of baseloaded electricity, as well as cogenerated electricity and process steam. In each case the HTGR-SC system has been evaluated against appropriate competing technologies. The computer code which was developed for this evaluation can be used to present the analyses on a cost of production or cash flow basis; thereby, presenting consistent results to a utility, interested in production costs, or an industrial steam user or third party investor, interested in returns on equity. Basically, there are two economic evaluation methodologies which can be used in the analysis of a project: (1) minimum revenue requirements, and (2) discounted cash flow

  8. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  9. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  10. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  11. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  12. Japanese HTTR program for demonstration of high temperature applications of nuclear energy

    International Nuclear Information System (INIS)

    Nishihara, T.; Hada, K.; Shiozawa, S.

    1997-01-01

    Construction works of the HTTR started in March 1991 in order to establish and upgrade the HTGR technology basis, to carry out innovative basic researches on high temperature engineering and to demonstrate high temperature heat utilization and application of nuclear heat. This report describes the demonstration program of high temperature heat utilization and application. (author). 2 refs, 4 figs, 3 tabs

  13. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  14. Concept of a Prestressed Cast Iron Pressure Vessel for a Modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Steinwarz, Wolfgang; Bounin, Dieter

    2014-01-01

    High Temperature Reactors (HTR) are representing one of the most interesting solutions for the upcoming generation of nuclear technology, especially with view to their inherent safety characteristics. To complete the safety concept of such plants already in the first phase of the technical development, Prestressed Cast Iron Pressure Vessels (PCIV) instead of the established forged steel reactor pressure vessels have been considered under the aspect of safety against bursting. A longterm research and development work, mainly performed in Germany, showed the excellent features of this technical solution. Diverse prototypic vessels were tested and officially proven. Design studies confirmed the feasibility of such a vessel concept also for Light Water Reactor types, too. The main concept elements of such a burst-proof vessel are: Strength and tightness functions are structurally separated. The tensile forces are carried by the prestressing systems consisting of a large number of independent wires. Compressive forces are applied to the vessel walls and heads. These are segmented into blocks of ductile cast iron. All cast iron blocks are prestressed to high levels of compression. The sealing function is assigned to a steel liner fixed to the cast iron blocks. The prestressing system is designed for an ultimate pressure of 2.3 times the design pressure. The prestress of the lids is designed for gapping at a much smaller pressure. Therefore, a drop of pressure will always occur before loss of strength (“leakage before failure”). In addition to these safety features further technical as well as economic aspects generate favorable assessment criteria: high design flexibility, feasibility of large vessel diameters; advantageous conditions for transport, assembly and decommissioning due to the segmented construction; advantage of workshop manufacturing; high-level quality control of components. Nowadays, considering the globally newly standardized safety requirements

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  17. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  18. Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

    International Nuclear Information System (INIS)

    Gokhan Yesilyurt; Hassan, Y.A.

    2003-01-01

    A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

  19. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  20. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  1. Development program for the high-temperature nuclear process heat system

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.

    1975-09-01

    A comprehensive development program plan for a high-temperature nuclear process heat system with a very high temperature gas-cooled reactor heat source is presented. The system would provide an interim substitute for fossil-fired sources and ultimately the vehicle for the production of substitute and synthetic fuels to replace petroleum and natural gas. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system has significant potential in a unique combination of the two sources that is environmentally and economically attractive and technically sound: the production of synthetic fuels from coal. In the longer term, it could be the key component in hydrogen production from water processes that offer a substitute fuel and chemical feedstock free of dependence on fossil-fuel reserves. The proposed development program is threefold: a process studies program, a demonstration plant program, and a supportive research and development program. Optional development scenarios are presented and evaluated, and a selection is proposed and qualified. The interdependence of the three major program elements is examined, but particular emphasis is placed on the supportive research and development activities. A detailed description of proposed activities in the supportive research and development program with tentative costs and schedules is presented as an appendix with an assessment of current status and planning

  2. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  3. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  4. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Lako, P.

    1996-06-01

    In this study the economic feasibility of new concepts of the High Temperature Reactor were investigated. These new concepts are characterized as inherently safe. The different concepts were used as industrial heat/power reactors and compared with a gas fired Steam and Gas turbine installation. The best economic advantages are offered by a HTR with a Thorium/Uranium cycle as compared with a gas fired steam- and gas turbine. 6 figs, 9 tabs, 21 refs

  5. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  6. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  8. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  9. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  10. IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: The DRAGON Reactor Experiment (DRE): The first demonstration High temperature gas reactor (HTGR) was built in the 1960's. Thirteen OECD countries began a project in 1959 to build an experimental reactor known as Dragon at Winfrith in the UK. The reactor - which operated successfully between 1966 and 1975 - had a thermal output of 20 MW and achieved a gas outlet temperature of 750 deg. C. The High Temperature Reactor concept, if it justified its expectations, was seen as having its place as an advanced thermal reactor between the current thermal reactor types such as the PWR, BWR, and AGR and the sodium cooled fast breeder reactor. It was expected that the HTR would offer better thermal efficiency, better uranium utilisation, either with low enriched uranium fuel or high enriched uranium thorium fuel, better inherent safety and lower unit power costs. In the event all these potential advantages were demonstrated to be in principle achievable. This view is still shared today. In fact Very High Temperature Reactors is one of the concepts retained for Generation IV. Projects on constructing Modular Pebble Bed Reactors are under way. Here all available Dragon Project Reports (DPR) - approximately 1000 - are collected in electronic form. An index points to the reports (PDF format); each table in the report is accessible in EXCEL format with the aim of facilitating access to the data. These reports describe the design, experiments and modelling carried out over a period of 17 years. 2 - Related or auxiliary information: IRPHE-HTR-ARCH-01, Archive of HTR Primary Documents NEA-1728/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  11. NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stevenson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsai, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating.

  12. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable

  13. Technology development for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Homan, F.J.; Turner, R.F.

    1989-01-01

    In the USA the Modular High-Temperature Gas-Cooled Reactor is in an advanced stage of design. The related HTGR program areas, the approaches to these programs along with sample results and a description of how these data are used are highlighted in the paper. (author). Figs and tabs

  14. Thermal insulation of the high-temperature helium-cooled reactors

    International Nuclear Information System (INIS)

    Kharlamov, A.G.; Grebennik, V.N.

    1979-01-01

    Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper

  15. Assessment of the high temperature fission chamber technology for the French fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l' Energie Atomique, CEA (France)

    2011-07-01

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  16. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)