WorldWideScience

Sample records for nuclear high temperature

  1. 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)

  2. 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

  3. 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.

  4. 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

  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. 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.

  7. 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

  8. 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

  9. 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)

  10. 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.)

  11. 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.

  12. 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

  13. 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)

  14. 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

  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. 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

  17. 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

  18. 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.

  19. 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)

  20. 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)

  1. Preliminary study on high temperature heat exchanger for nuclear steel making

    Energy Technology Data Exchange (ETDEWEB)

    Nakada, T; Ohtomo, A; Yamada, R; Suzuki, K; Narita, Y [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

    1975-05-01

    Both in the high temperature heat exchanger and in the steam reformer, there remain several technical problems to be solved before nuclear steel making is actualized. The loop for use with basic studies of those problems was planned by the Iron and Steel Institute of Japan (ISIJ), and its actual design, construction and co-ordination of tests were undertaken by IHI on behalf of ISIJ. The primary coolant used in the loop was helium having a pressure of approx. 12 kg/cm/sup 2/g and a temperature of approx. 1100/sup 0/C at the inlet of the high temperature heat exchanger, i.e., the test section. Steam, hydrogen, and carbon monoxide were used as secondary coolants. Of the technical problems regarding the high temperature heat exchanger for nuclear steel making, which were selected and studied using the loop, the following items are discussed: (1) heat exchange performance using helium and steam; (2) hydrogen permeation of heat resisting alloys; (3) creep and carburization of heat resisting alloys; amd (4) hydrogen absorption performance of the titanium sponge.

  2. Application on electrochemistry measurement of high temperature high pressure condition in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Li Yuchun; Xiao Zhongliang; Jiang Ya; Yu Xiaowei; Pang Feifei; Deng Fenfang; Gao Fan; Zhou Nianguang

    2011-01-01

    High temperature high pressure electrochemistry testing system was comprehensively analyzed in this paper, according to actual status for supervision in primary and secondary circuits of PWR nuclear power plants. Three research methods were reviewed and discussed for in-situ monitor system. By combination with ECP realtime measurement it was executed for evaluation and water chemistry optimization in nuclear power plants. It is pointed out that in-situ electrochemistry measurement has great potential application for water chemistry evaluation in PWR nuclear power plants. (authors)

  3. 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.

  4. Design and evaluation of a pressure sensor for high temperature nuclear application

    International Nuclear Information System (INIS)

    Yancey, M.E.

    1981-11-01

    The goal of this technical development task was the development of a small eddy-current pressure sensor for use within a high temperature nuclear environment. The sensor is designed for use at pressures and temperatures of up to 17.23 MPa and 650 0 F. The design of the sensor incorporated features to minimize possible errors due to temperature transients present in nuclear applications. This report describes a prototype pressure sensor that was designed, the associated 100 kHz signal conditioning electronics, and the evaluation tests which were conducted

  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. 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

  7. 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

  8. 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

  9. Assessment of ambient-temperature, high-resolution detectors for nuclear safeguards applications

    International Nuclear Information System (INIS)

    Ruhter, W.D.; McQuaid, J.H.; Lavietes, A.

    1993-01-01

    High-resolution, gamma- and x-ray spectrometry are used routinely in nuclear safeguards verification measurements of plutonium and uranium in the field. These measurements are now performed with high-purity germanium (HPGe) detectors that require cooling liquid-nitrogen temperatures, thus limiting their utility in field and unattended safeguards measurement applications. Ambient temperature semiconductor detectors may complement HPGe detectors for certain safeguards verification applications. Their potential will be determined by criteria such as their performance, commercial availability, stage of development, and costs. We have conducted as assessment of ambient temperature detectors for safeguards measurement applications with these criteria in mind

  10. Lightweight Damage Tolerant, High-Temperature Radiators for Nuclear Power and Propulsion

    Science.gov (United States)

    Craven, Paul D.; SanSoucie, Michael P.

    2015-01-01

    NASA is increasingly emphasizing exploration to bodies beyond near-Earth orbit. New propulsion systems and new spacecraft are being built for these missions. As the target bodies get further out from Earth, high energy density systems, e.g., nuclear fusion, for propulsion and power will be advantageous. The mass and size of these systems, including supporting systems such as the heat exchange system, including thermal radiators, will need to be as small as possible. Conventional heat exchange systems are a significant portion of the total thermal management mass and size. Nuclear electric propulsion (NEP) is a promising option for high-speed, in-space travel due to the high energy density of nuclear fission power sources and efficient electric thrusters. Heat from the reactor is converted to power for use in propulsion or for system power. The heat not used in the power conversion is then radiated to space as shown in figure 1. Advanced power conversion technologies will require high operating temperatures and would benefit from lightweight radiator materials. Radiator performance dictates power output for nuclear electric propulsion systems. Pitch-based carbon fiber materials have the potential to offer significant improvements in operating temperature, thermal conductivity, and mass. These properties combine to allow significant decreases in the total mass of the radiators and significant increases in the operating temperature of the fins. A Center-funded project at NASA Marshall Space Flight Center has shown that high thermal conductivity, woven carbon fiber fins with no matrix material, can be used to dissipate waste heat from NEP systems and because of high specific power (kW/kg), will require less mass and possibly less total area than standard metal and composite radiator fins for radiating the same amount of heat. This project uses an innovative approach to reduce the mass and size required for the thermal radiators to the point that in-space NEP and power

  11. 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.)

  12. Preliminary study on high temperature heat exchanger for nuclear steel making

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Y [Tokyo Inst. of Tech. (Japan); Ikegami, H

    1975-03-01

    In the high temperature heat exchanger as well as the steam reformer, several technical problems should be solved before realizing a nuclear plant complex for iron and steel making. Research has been carried out on heat exchanger between helium and steam, hydrogen permeation through super alloys, hydrogen removal using a titanium sponge, and creep and carburization performance of super alloys. The primary coolant used is helium having a pressure of approximately 12 kg/cm/sup 2/G and a temperature of approximately 1100/sup 0/C measured at the inlet of the high temperature heat exchanger, i.e., the test section. Steam, hydrogen and carbon monoxide are used as secondary coolants.

  13. 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)

  14. Muon nuclear fusion and low temperature nuclear fusion

    International Nuclear Information System (INIS)

    Nagamine, Kanetada

    1990-01-01

    Low temperature (or normal temperature) nuclear fusion is one of the phenomena causing nuclear fusion without requiring high temperature. In thermal nuclear fusion, the Coulomb barrier is overcome with the help of thermal energy, but in the low temperature nuclear fusion, the Coulomb barrier is neutralized by the introduction of the particles having larger mass than electrons and negative charges, at this time, if two nuclei can approach to the distance of 10 -13 cm in the neutral state, the occurrence of nuclear fusion reaction is expected. As the mass of the particles is heavier, the neutral region is smaller, and nuclear fusion is easy to occur. The particles to meet this purpose are the electrons within substances and muons. The research on muon nuclear fusion became suddenly active in the latter half of 1970s, the cause of which was the discovery of the fact that the formation of muons occurs resonantly rapidly in D-T and D-D systems. Muons are the unstable elementary particles having the life of 2.2 μs, and they can have positive and negative charges. In the muon catalyzed fusion, the muons with negative charge take part. The principle of the muon catalyzed fusion, its present status and future perspective, and the present status of low temperature nuclear fusion are reported. (K.I.)

  15. The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

  16. 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

  17. 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.

  18. 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)

  19. 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

  20. 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.)

  1. Friction and wear studies of nuclear power plant components in pressurized high temperature water environments

    International Nuclear Information System (INIS)

    Ko, P.L.; Zbinden, M.; Taponat, M.C.; Robertson, M.F.

    1997-01-01

    The present paper is part of a series of papers aiming to present the friction and wear results of a collaborative study on nuclear power plant components tested in pressurized high temperature water. The high temperature test facilities and the methodology in presenting the kinetics and wear results are described in detail. The results of the same material combinations obtained from two very different high temperature test facilities (NRCC and EDF) are presented and discussed. (K.A.)

  2. 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.)

  3. 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.)

  4. High temperature materials

    International Nuclear Information System (INIS)

    2003-01-01

    The aim of this workshop is to share the needs of high temperature and nuclear fuel materials for future nuclear systems, to take stock of the status of researches in this domain and to propose some cooperation works between the different research organisations. The future nuclear systems are the very high temperature (850 to 1200 deg. C) gas cooled reactors (GCR) and the molten salt reactors (MSR). These systems include not only the reactor but also the fabrication and reprocessing of the spent fuel. This document brings together the transparencies of 13 communications among the 25 given at the workshop: 1) characteristics and needs of future systems: specifications, materials and fuel needs for fast spectrum GCR and very high temperature GCR; 2) high temperature materials out of neutron flux: thermal barriers: materials, resistance, lifetimes; nickel-base metal alloys: status of knowledge, mechanical behaviour, possible applications; corrosion linked with the gas coolant: knowledge and problems to be solved; super-alloys for turbines: alloys for blades and discs; corrosion linked with MSR: knowledge and problems to be solved; 3) materials for reactor core structure: nuclear graphite and carbon; fuel assembly structure materials of the GCR with fast neutron spectrum: status of knowledge and ceramics and cermets needs; silicon carbide as fuel confinement material, study of irradiation induced defects; migration of fission products, I and Cs in SiC; 4) materials for hydrogen production: status of the knowledge and needs for the thermochemical cycle; 5) technologies: GCR components and the associated material needs: compact exchangers, pumps, turbines; MSR components: valves, exchangers, pumps. (J.S.)

  5. Novel Methods of Tritium Sequestration: High Temperature Gettering and Separation Membrane Materials Discovery for Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Franglin [Univ. of South Carolina, Columbia, SC (United States); Sholl, David [Georgia Inst. of Technology, Atlanta, GA (United States); Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Lyer, Ratnasabapathy [Claflin Univ., Orangeburg, SC (United States); Iyer, Ratnasabapathy [Claflin Univ., Orangeburg, SC (United States); Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States)

    2015-01-22

    This project is aimed at addressing critical issues related to tritium sequestration in next generation nuclear energy systems. A technical hurdle to the use of high temperature heat from the exhaust produced in the next generation nuclear processes in commercial applications such as nuclear hydrogen production is the trace level of tritium present in the exhaust gas streams. This presents a significant challenge since the removal of tritium from the high temperature gas stream must be accomplished at elevated temperatures in order to subsequently make use of this heat in downstream processing. One aspect of the current project is to extend the techniques and knowledge base for metal hydride materials being developed for the ''hydrogen economy'' based on low temperature absorption/desorption of hydrogen to develop materials with adequate thermal stability and an affinity for hydrogen at elevated temperatures. The second focus area of this project is to evaluate high temperature proton conducting materials as hydrogen isotope separation membranes. Both computational and experimental approaches will be applied to enhance the knowledge base of hydrogen interactions with metal and metal oxide materials. The common theme between both branches of research is the emphasis on both composition and microstructure influence on the performance of sequestration materials.

  6. Novel Methods of Tritium Sequestration: High Temperature Gettering and Separation Membrane Materials Discovery for Nuclear Energy Systems

    International Nuclear Information System (INIS)

    2015-01-01

    This project is aimed at addressing critical issues related to tritium sequestration in next generation nuclear energy systems. A technical hurdle to the use of high temperature heat from the exhaust produced in the next generation nuclear processes in commercial applications such as nuclear hydrogen production is the trace level of tritium present in the exhaust gas streams. This presents a significant challenge since the removal of tritium from the high temperature gas stream must be accomplished at elevated temperatures in order to subsequently make use of this heat in downstream processing. One aspect of the current project is to extend the techniques and knowledge base for metal hydride materials being developed for the ''hydrogen economy'' based on low temperature absorption/desorption of hydrogen to develop materials with adequate thermal stability and an affinity for hydrogen at elevated temperatures. The second focus area of this project is to evaluate high temperature proton conducting materials as hydrogen isotope separation membranes. Both computational and experimental approaches will be applied to enhance the knowledge base of hydrogen interactions with metal and metal oxide materials. The common theme between both branches of research is the emphasis on both composition and microstructure influence on the performance of sequestration materials.

  7. 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

  8. Study on microstructure and high temperature wear resistance of laser cladded nuclear valve clack

    International Nuclear Information System (INIS)

    Zhang Chunliang; Chen Zichen

    2002-01-01

    Laser cladding of Co-base alloy on the nuclear valve-sealing surface are performed with a 5 kW CO 2 transverse flowing laser. The microstructure and the high temperature impact-slide wear resistance of the laser cladded coating and the plasma cladded coating are studied. The results show that the microstructure, the dilution rate and the high temperature impact-slide wear resistance of the laser cladded coating have obvious advantages over the spurt cladding processing

  9. 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)

  10. Magnetostrictive device for high-temperature sound and vibration measurement in nuclear power stations

    International Nuclear Information System (INIS)

    Hans, R.; Podgorski, J.

    1977-01-01

    The demands on the monitoring systems in nuclear power stations are increasing continuously, not only because of more stringent safety requirements but also for reasons of plant availability and thus economic efficiency. The noise and vibration measurements which therefore have to be taken make it necessary to provide measuring devices with a high degree of efficiency, adequate sensitivity and resistance to high temperatures, radiation and corrosion. Probes using the magnetostrictive effect, whereby a ferromagnetic core changes its length in a magnetic field - a phenomenon which has been known for approximately fifty years - fulfill all the conditions for application in nuclear power stations. (orig.) [de

  11. High temperature materials; Materiaux a hautes temperatures

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    The aim of this workshop is to share the needs of high temperature and nuclear fuel materials for future nuclear systems, to take stock of the status of researches in this domain and to propose some cooperation works between the different research organisations. The future nuclear systems are the very high temperature (850 to 1200 deg. C) gas cooled reactors (GCR) and the molten salt reactors (MSR). These systems include not only the reactor but also the fabrication and reprocessing of the spent fuel. This document brings together the transparencies of 13 communications among the 25 given at the workshop: 1) characteristics and needs of future systems: specifications, materials and fuel needs for fast spectrum GCR and very high temperature GCR; 2) high temperature materials out of neutron flux: thermal barriers: materials, resistance, lifetimes; nickel-base metal alloys: status of knowledge, mechanical behaviour, possible applications; corrosion linked with the gas coolant: knowledge and problems to be solved; super-alloys for turbines: alloys for blades and discs; corrosion linked with MSR: knowledge and problems to be solved; 3) materials for reactor core structure: nuclear graphite and carbon; fuel assembly structure materials of the GCR with fast neutron spectrum: status of knowledge and ceramics and cermets needs; silicon carbide as fuel confinement material, study of irradiation induced defects; migration of fission products, I and Cs in SiC; 4) materials for hydrogen production: status of the knowledge and needs for the thermochemical cycle; 5) technologies: GCR components and the associated material needs: compact exchangers, pumps, turbines; MSR components: valves, exchangers, pumps. (J.S.)

  12. 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

  13. 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)

  14. 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

  15. Development of coating technology for nuclear fuel by self-propagating high temperature synthesis

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, Bong G.; Lee, Y. W.

    1997-01-01

    This paper presents experimental results of the preparation of silicon carbide and graphite layers on a nuclear fuel from silane and propane gases by a conventional chemical vapor deposition and combustion synthesis technologies. The direct reaction between silicon and pyrolytic carbon in a high temperature releases sufficient amount of energy to make a synthesis self-sustaining under the preheating of about 1200 deg C. During this high temperature process, lamellar structure with isotropic carbon synthesis. A full characterization of phase composition and final morphology of the coated layers by X-ray diffraction, SEM and AES is presented. (author). 6 refs., 1 tab., 11 figs

  16. 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)

  17. Irradiation effects on C/C composite materials for high temperature nuclear applications

    International Nuclear Information System (INIS)

    Eto, M.; Ugachi, H.; Baba, S.I.; Ishiyama, S.; Ishihara, M.; Hayashi, K.

    2000-01-01

    Excellent characteristics such as high strength and high thermal shock resistance of C/C composite materials have led us to try to apply them to the high temperature components in nuclear facilities. Such components include the armour tile of the first wall and divertor of fusion reactor and the elements of control rod for the use in HTGR. One of the most important aspects to be clarified about C/C composites for nuclear applications is the effect of neutron irradiation on their properties. At the Japan Atomic Energy Research Institute (JAERI), research on the irradiation effects on various properties of C/C composite materials has been carried out using fission reactors (JRR-3, JMTR), accelerators (TANDEM, TIARA) and the Fusion Neutronics Source (FNS). Additionally, strength tests of some neutron-irradiated elements for the control rod were carried out to investigate the feasibility of C/C composites. The paper summarises the R and D activities on the irradiation effects on C/C composites. (authors)

  18. 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.

  19. Development of High Temperature Chemistry Measurement System for Establishment of On-Line Water Chemistry Surveillance Network in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, Won Ho; Song, Kyu Seok; Joo, Ki Soo; Choi, Ke Chon; Ha, Yeong Keong; Ahn, Hong Joo; Im, Hee Jung; Maeng, Wan Young

    2010-07-01

    An integrated high-temperature water chemistry sensor (pH, E redox ) was developed for the establishment of the on-line water chemistry surveillance system in nuclear power plants. The basic performance of the integrated sensor was confirmed in high-temperature (280 .deg. C, 150kg/m 2 ) lithium borate solutions by using the relationship between the concentration of lithium ion and pH-E redox values. Especially, the effects of various environmental factors such as temperature, pressure, and flow rate on YSZ-based pH electrode were evaluated for ensuring the accuracy of high-temperature pH measurement. And the relationships between each water chemistry factor (pH, redox potential, electrical conductivity) were induced for enhancing the credibility of water chemistry measurement. In addition, on the basis of the evaluation of a nuclear plant design company, we suggested potential installation positions of the measurement system in a nuclear power plant

  20. 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.

  1. 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)

  2. 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)

  3. 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.)

  4. 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%

  5. 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

  6. 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)

  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. 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

  9. 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

  10. 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.

  11. Study on extreme high temperature of cooling water in Chinese coastal nuclear power plant

    International Nuclear Information System (INIS)

    Yu Fan; Jiang Ziying

    2012-01-01

    In order to protect aquatic life from the harmful effects of thermal discharge, the appropriate water temperature limits or the scope of the mixing zone is a key issue in the regulatory control of the environmental impact of thermal discharge. Based on the sea surface temperature in the Chinese coastal waters, the extreme value of the seawater temperature change was analyzed by using the Gumbel model. The limit of the design temperature rise of cooling water in the outfall is 9 ℃, and the limit of the temperature rise of cooling water in the edge of the mixing zone is 4 ℃. The extreme high temperature of the cooling water in Chinese coastal nuclear power plant is 37 ℃ in the Bohai Sea, Yellow Sea, and is 40 ℃ in East China Sea, South China Sea. (authors)

  12. Temperature measuring element in nuclear reactors

    International Nuclear Information System (INIS)

    Wada, Takashi.

    1987-01-01

    Purpose: To easily measure the partial maximum temperature at a portion within the nuclear reactor where the connection with the external portion is difficult. Constitution: Sodium, potassium or the alloy thereof with high heat expansion coefficient is packed into an elastic vessel having elastic walls contained in a capsule. A piercing member formed into an acute triangle is attached to one end in the direction of expansion and contraction of the elastic container. The two sides of the triangle form an acute knife edge. A diaphragm is disposed within a capsule at a position opposed to the sharpened direction of the piercing member. Such a capsule is placed in a predetermined position of the nuclear reactor. The elastic vessel causes thermal expansion displacement depending on the temperature at a certain position, by which the top end of the pierce member penetrates through the diaphragm. A pierced scar of a penetration length depending on the temperature is resulted to the diaphragm. The length of the piercing damage is electroscopically observed and compared with the calibration curve to recognize the maximum temperature in the predetermined portion of the nuclear reactor. (Kamimura, M.)

  13. Ultra-high temperature direct propulsion

    International Nuclear Information System (INIS)

    Araj, K.J.; Slovik, G.; Powell, J.R.; Ludewig, H.

    1987-01-01

    Potential advantages of ultra-high exhaust temperature (3000 K - 4000 K) direct propulsion nuclear rockets are explored. Modifications to the Particle Bed Reactor (PBR) to achieve these temperatures are described. Benefits of ultra-high temperature propulsion are discussed for two missions - orbit transfer (ΔV = 5546 m/s) and interplanetary exploration (ΔV = 20000 m/s). For such missions ultra-high temperatures appear to be worth the additional complexity. Thrust levels are reduced substantially for a given power level, due to the higher enthalpy caused by partial disassociation of the hydrogen propellant. Though technically challenging, it appears potentially feasible to achieve such ultra high temperatures using the PBR

  14. 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

  15. 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

  16. Spin temperature concept verified by optical magnetometry of nuclear spins

    Science.gov (United States)

    Vladimirova, M.; Cronenberger, S.; Scalbert, D.; Ryzhov, I. I.; Zapasskii, V. S.; Kozlov, G. G.; Lemaître, A.; Kavokin, K. V.

    2018-01-01

    We develop a method of nonperturbative optical control over adiabatic remagnetization of the nuclear spin system and apply it to verify the spin temperature concept in GaAs microcavities. The nuclear spin system is shown to exactly follow the predictions of the spin temperature theory, despite the quadrupole interaction that was earlier reported to disrupt nuclear spin thermalization. These findings open a way for the deep cooling of nuclear spins in semiconductor structures, with the prospect of realizing nuclear spin-ordered states for high-fidelity spin-photon interfaces.

  17. 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

  18. High Temperature Electrolysis for Hydrogen Production from Nuclear Energy - Technology Summary

    International Nuclear Information System (INIS)

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

    2010-01-01

    The Department of Energy, Office of Nuclear Energy, has requested that a Hydrogen Technology Down-Selection be performed to identify the hydrogen production technology that has the best potential for timely commercial demonstration and for ultimate deployment with the Next Generation Nuclear Plant (NGNP). An Independent Review Team has been assembled to execute the down-selection. This report has been prepared to provide the members of the Independent Review Team with detailed background information on the High Temperature Electrolysis (HTE) process, hardware, and state of the art. The Idaho National Laboratory has been serving as the lead lab for HTE research and development under the Nuclear Hydrogen Initiative. The INL HTE program has included small-scale experiments, detailed computational modeling, system modeling, and technology demonstration. Aspects of all of these activities are included in this report. In terms of technology demonstration, the INL successfully completed a 1000-hour test of the HTE Integrated Laboratory Scale (ILS) technology demonstration experiment during the fall of 2008. The HTE ILS achieved a hydrogen production rate in excess of 5.7 Nm3/hr, with a power consumption of 18 kW. This hydrogen production rate is far larger than has been demonstrated by any of the thermochemical or hybrid processes to date.

  19. 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

  20. 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

  1. High temperature corrosion in the service environments of a nuclear process heat plant

    International Nuclear Information System (INIS)

    Quadakkers, W.J.

    1987-01-01

    In a nuclear process heat plant the heat-exchanging components fabricated from nickel- and Fe-Ni-based alloys are subjected to corrosive service environments at temperatures up to 950 0 C for service lives of up to 140 000 h. In this paper the corrosion behaviour of the high temperature alloys in the different service environments will be described. It is shown that the degree of protection provided by Cr 2 O 3 -based surface oxide scales against carburization and decarburization of the alloys is primarily determined not by the oxidation potential of the atmospheres but by a dynamic process involving, on the one hand, the oxidizing gas species and the metal and, on the other hand, the carbon in the alloy and the oxide scale. (orig.)

  2. A dilution refrigerator combining low base temperature, high cooling power and low heat leak for use with nuclear cooling

    International Nuclear Information System (INIS)

    Bradley, D.I.; Guenault, A.M.; Keith, V.; Miller, I.E.; Pickett, G.R.; Bradshaw, T.W.; Locke-Scobie, B.G.

    1982-01-01

    The design philosophy, design, construction and performance of a dilution refrigerator specifically intended for nuclear cooling experiments in the submillikelvin regime is described. Attention has been paid from the outset to minimizing sources of heat leaks, and to achieving a low base temperature and relatively high cooling power below 10 mK. The refrigerator uses sintered silver heat exchangers similar to those developed at Grenoble. The machine has a base temperature of 3 mK or lower and can precool a copper nuclear specimen in 6.8 T to 8 mK in 70 h. The heat leak to the innermost nuclear stage is < 30 pW after only a few days' running. (author)

  3. Future needs for inelastic analysis in design of high-temperature nuclear plant components

    International Nuclear Information System (INIS)

    Corum, J.M.

    1980-01-01

    The role that inelastic analyses play in the design of high-temperature nuclear plant components is described. The design methodology, which explicitly accounts for nonlinear material deformation and time-dependent failure modes, requires a significant level of realism in the prediction of structural response. Thus, material deformation and failure modeling are, along with computational procedures, key parts of the methodology. Each of these is briefly discussed along with validation by comparisons with benchmark structural tests, and problem areas and needs are discussed for each

  4. High-temperature deformation and processing maps of Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles

    Science.gov (United States)

    Chen, Jing; Liu, Huiqun; Zhang, Ruiqian; Li, Gang; Yi, Danqing; Lin, Gaoyong; Guo, Zhen; Liu, Shaoqiang

    2018-06-01

    High-temperature compression deformation of a Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles was investigated at 750 °C-950 °C with a strain rate of 0.01-1.0 s-1 and height reduction of 20%. Scanning electron microscopy was utilized to investigate the influence of the deformation conditions on the microstructure of the composite and damage to the coated surrogate fuel particles. The results indicated that the flow stress of the composite increased with increasing strain rate and decreasing temperature. The true stress-strain curves showed obvious serrated oscillation characteristics. There were stable deformation ranges at the initial deformation stage with low true strain at strain rate 0.01 s-1 for all measured temperatures. Additionally, the coating on the surface of the surrogate nuclear fuel particles was damaged when the Zr-4 matrix was deformed at conditions of high strain rate and low temperature. The deformation stability was obtained from the processing maps and microstructural characterization. The high-temperature deformation activation energy was 354.22, 407.68, and 433.81 kJ/mol at true strains of 0.02, 0.08, and 0.15, respectively. The optimum deformation parameters for the composite were 900-950 °C and 0.01 s-1. These results are expected to provide guidance for subsequent determination of possible hot working processes for this composite.

  5. 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.

  6. 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

  7. 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)

  8. 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)

  9. High Temperature Electrolysis for Hydrogen Production from Nuclear Energy – TechnologySummary

    Energy Technology Data Exchange (ETDEWEB)

    J. E. O' Brien; C. M. Stoots; J. S. Herring; M. G. McKellar; E. A. Harvego; M. S. Sohal; K. G. Condie

    2010-02-01

    The Department of Energy, Office of Nuclear Energy, has requested that a Hydrogen Technology Down-Selection be performed to identify the hydrogen production technology that has the best potential for timely commercial demonstration and for ultimate deployment with the Next Generation Nuclear Plant (NGNP). An Independent Review Team has been assembled to execute the down-selection. This report has been prepared to provide the members of the Independent Review Team with detailed background information on the High Temperature Electrolysis (HTE) process, hardware, and state of the art. The Idaho National Laboratory has been serving as the lead lab for HTE research and development under the Nuclear Hydrogen Initiative. The INL HTE program has included small-scale experiments, detailed computational modeling, system modeling, and technology demonstration. Aspects of all of these activities are included in this report. In terms of technology demonstration, the INL successfully completed a 1000-hour test of the HTE Integrated Laboratory Scale (ILS) technology demonstration experiment during the fall of 2008. The HTE ILS achieved a hydrogen production rate in excess of 5.7 Nm3/hr, with a power consumption of 18 kW. This hydrogen production rate is far larger than has been demonstrated by any of the thermochemical or hybrid processes to date.

  10. Evolution of nuclear collectivity at high spins and temperatures

    International Nuclear Information System (INIS)

    Baktash, C.

    1989-01-01

    In the past few years, we have utilized the Spin Spectrometer and a variety of complementary probes (continuum γrays, proton-γ coincidence spectroscopy and γ decay of GDR) to study the nuclear response to the DIFFERENTIAL effects of increasing spin and temperature for constant values of excitation energy or spin, respectively. In this paper we shall describe two of the experiments that trace the properties of rapidly-rotating nuclei at small to moderate excitation energies. 22 refs., 7 figs

  11. Study on the Simulation of Crud Formation using Piping Materials of Nuclear Power Plant in High Temperature Water

    International Nuclear Information System (INIS)

    Kim, Sang Hyun; Kim, In Sup; Lee, Kun Jai

    2005-01-01

    High temperature - high pressure apparatus was developed to simulate nickel fewite corrosion products which were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needle like or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.

  12. 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.)

  13. Design methods for high temperature power plant structures

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1984-01-01

    The subject is discussed under the headings: introduction (scope of paper - reviews of design methods and design criteria currently in use for both nuclear and fossil fuelled power plant; examples chosen are (a) BS 1113, representative of design codes employed for power station boiler plant; (b) ASME Code Case N47, which is being developed for high temperature nuclear reactors, especially the liquid metal fast breeder reactor); design codes for power station boilers; Code Case N47 (design in the absence of thermal shock and thermal fatigue; design against cyclic loading at high temperature; further research in support of high temperature design methods and criteria for LMFBRs); concluding remarks. (U.K.)

  14. 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

  15. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, G.; Radecka, A.; Massey, C.

    2015-01-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  16. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

    2015-07-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  17. 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.

  18. A high-throughput investigation of Fe-Cr-Al as a novel high-temperature coating for nuclear cladding materials.

    Science.gov (United States)

    Bunn, Jonathan Kenneth; Fang, Randy L; Albing, Mark R; Mehta, Apurva; Kramer, Matthew J; Besser, Matthew F; Hattrick-Simpers, Jason R

    2015-07-10

    High-temperature alloy coatings that can resist oxidation are urgently needed as nuclear cladding materials to mitigate the danger of hydrogen explosions during meltdown. Here we apply a combination of computationally guided materials synthesis, high-throughput structural characterization and data analysis tools to investigate the feasibility of coatings from the Fe–Cr–Al alloy system. Composition-spread samples were synthesized to cover the region of the phase diagram previous bulk studies have identified as forming protective oxides. The metallurgical and oxide phase evolution were studied via in situ synchrotron glancing incidence x-ray diffraction at temperatures up to 690 K. A composition region with an Al concentration greater than 3.08 at%, and between 20.0 at% and 32.9 at% Cr showed the least overall oxide growth. Subsequently, a series of samples were deposited on stubs and their oxidation behavior at 1373 K was observed. The continued presence of a passivating oxide was confirmed in this region over a period of 6 h.

  19. High-temperature materials and structural ceramics

    International Nuclear Information System (INIS)

    1990-01-01

    This report gives a survey of research work in the area of high-temperature materials and structural ceramics of the KFA (Juelich Nuclear Research Center). The following topics are treated: (1) For energy facilities: ODS materials for gas turbine blades and heat exchangers; assessment of the remaining life of main steam pipes, material characterization and material stress limits for First-Wall components; metallic and graphitic materials for high-temperature reactors. (2) For process engineering plants: composites for reformer tubes and cracking tubes; ceramic/ceramic joints and metal/ceramic and metal/metal joints; Composites and alloys for rolling bearing and sliding systems up to application temperatures of 1000deg C; high-temperature corrosion of metal and ceramic material; porous ceramic high-temperature filters and moulding coat-mix techniques; electrically conducting ceramic material (superconductors, fuel cells, solid electrolytes); high-temperature light sources (high-temperature chemistry); oil vapor engines with caramic components; ODS materials for components in diesel engines and vehicle gas turbines. (MM) [de

  20. 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

  1. Materials corrosion and protection at high temperatures

    International Nuclear Information System (INIS)

    Balbaud, F.; Desgranges, Clara; Martinelli, Laure; Rouillard, Fabien; Duhamel, Cecile; Marchetti, Loic; Perrin, Stephane; Molins, Regine; Chevalier, S.; Heintz, O.; David, N.; Fiorani, J.M.; Vilasi, M.; Wouters, Y.; Galerie, A.; Mangelinck, D.; Viguier, B.; Monceau, D.; Soustelle, M.; Pijolat, M.; Favergeon, J.; Brancherie, D.; Moulin, G.; Dawi, K.; Wolski, K.; Barnier, V.; Rebillat, F.; Lavigne, O.; Brossard, J.M.; Ropital, F.; Mougin, J.

    2011-01-01

    This book was made from the lectures given in 2010 at the thematic school on 'materials corrosion and protection at high temperatures'. It gathers the contributions from scientists and engineers coming from various communities and presents a state-of-the-art of the scientific and technological developments concerning the behaviour of materials at high temperature, in aggressive environments and in various domains (aerospace, nuclear, energy valorization, and chemical industries). It supplies pedagogical tools to grasp high temperature corrosion thanks to the understanding of oxidation mechanisms. It proposes some protection solutions for materials and structures. Content: 1 - corrosion costs; macro-economical and metallurgical approach; 2 - basic concepts of thermo-chemistry; 3 - introduction to the Calphad (calculation of phase diagrams) method; 4 - use of the thermodynamic tool: application to pack-cementation; 5 - elements of crystallography and of real solids description; 6 - diffusion in solids; 7 - notions of mechanics inside crystals; 8 - high temperature corrosion: phenomena, models, simulations; 9 - pseudo-stationary regime in heterogeneous kinetics; 10 - nucleation, growth and kinetic models; 11 - test experiments in heterogeneous kinetics; 12 - mechanical aspects of metal/oxide systems; 13 - coupling phenomena in high temperature oxidation; 14 - other corrosion types; 15 - methods of oxidized surfaces analysis at micro- and nano-scales; 16 - use of SIMS in the study of high temperature corrosion of metals and alloys; 17 - oxidation of ceramics and of ceramic matrix composite materials; 18 - protective coatings against corrosion and oxidation; 19 - high temperature corrosion in the 4. generation of nuclear reactor systems; 20 - heat exchangers corrosion in municipal waste energy valorization facilities; 21 - high temperature corrosion in oil refining and petrochemistry; 22 - high temperature corrosion in new energies industry. (J.S.)

  2. 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%.

  3. Temperature dependence of nuclear surface properties

    International Nuclear Information System (INIS)

    Campi, X.; Stringari, S.

    1982-01-01

    Thermal properties of nuclear surface are investigated in a semi-infinite medium. Explicit analytical expression are given for the temperature dependence of surface thickness, surface energy and surface free energy. In this model the temperature effects depend critically on the nuclear incompressibility and on the shape of the effective mass at the surface. To illustrate the relevance of these effects we made an estimate of the temperature dependence of the fission barrier height. (orig.)

  4. 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.

  5. 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.

  6. 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

  7. Characterizing high-temperature deformation of internally heated nuclear fuel element simulators

    Energy Technology Data Exchange (ETDEWEB)

    Belov, A.I.; Fong, R.W.L.; Leitch, B.W.; Nitheanandan, T.; Williams, A., E-mail: alexander.belov@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 {sup o}C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 {sup o}C, only minor lateral deflections of the element were observed. Further heating to above 700 {sup o}C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a 'hard' pellet-cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down. (author)

  8. 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).

  9. 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

  10. 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

  11. 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)

  12. 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.

  13. 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.

  14. 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

  15. Generating highly polarized nuclear spins in solution using dynamic nuclear polarization

    DEFF Research Database (Denmark)

    Wolber, J.; Ellner, F.; Fridlund, B.

    2004-01-01

    A method to generate strongly polarized nuclear spins in solution has been developed, using Dynamic Nuclear Polarization (DNP) at a temperature of 1.2K, and at a field of 3.354T, corresponding to an electron spin resonance frequency of 94GHz. Trityl radicals are used to directly polarize 13C...... and other low-γ nuclei. Subsequent to the DNP process, the solid sample is dissolved rapidly with a warm solvent to create a solution of molecules with highly polarized nuclear spins. Two main applications are proposed: high-resolution liquid state NMR with enhanced sensitivity, and the use...

  16. Economic Analysis of the Reference Design for a Nuclear-Driven 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 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 consists of 4,009,177 cells with a per-cell active area of 225 cm2. A nominal cell area-specific resistance, ASR, value of 0.4 Ohm-cm2 with a current density of 0.25 A/cm2 was used, and isothermal boundary conditions were assumed. 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 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 is 96%. The overall system thermal-to-hydrogen production efficiency (based on the low heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of the plant was also performed using the H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program. 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 using realistic financial and cost estimating assumptions. A required cost of $3.23 per kg of hydrogen produced was calculated assuming an internal rate of return of 10%. Approximately 73% of this cost ($2.36/kg) is the result of capital costs associated with

  17. 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

  18. Advances in high temperature water chemistry and future issues

    International Nuclear Information System (INIS)

    Millett, P.J.

    2005-01-01

    This paper traces the development of advances in high temperature water chemistry with emphasis in the field of nuclear power. Many of the water chemistry technologies used in plants throughout the world today would not have been possible without the underlying scientific advances made in this field. In recent years, optimization of water chemistry has been accomplished by the availability of high temperature water chemistry codes such as MULTEQ. These tools have made the science of high temperature chemistry readily accessible for engineering purposes. The paper closes with a discussion of what additional scientific data and insights must be pursued in order to support the further development of water chemistry technologies for the nuclear industry. (orig.)

  19. Structural analysis technology for high-temperature design

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1977-01-01

    Results from an ongoing program devoted to the development of verified high-temperature structural design technology applicable to nuclear reactor systems are described. The major aspects addressed by the program are (1) deformation behavior; (2) failure associated with creep rupture, brittle fracture, fatigue, creep-fatigue interactions, and crack propagation; and (3) the establishment of appropriate design criteria. This paper discusses information developed in the deformation behavior category. The material considered is type 304 stainless steel, and the temperatures range to 1100 0 F (593 0 C). In essence, the paper considers the ingredients necessary for predicting relatively high-temperature inelastic deformation behavior of engineering structures under time-varying temperature and load conditions and gives some examples. These examples illustrate the utility and acceptability of the computational methods identified and developed for prediting essential features of complex inelastic behaviors. Conditions and responses that can be encountered under nuclear reactor service conditions and invoked in the examples. (Auth.)

  20. 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

  1. Elevated temperature and high pressure large helium gas loop

    International Nuclear Information System (INIS)

    Sakasai, Minoru; Midoriyama, Shigeru; Miyata, Toyohiko; Nakase, Tsuyoshi; Izaki, Makoto

    1979-01-01

    The development of high temperature gas-cooled reactors especially aiming at the multi-purpose utilization of nuclear heat energy is carried out actively in Japan and West Germany. In Japan, the experimental HTGR of 50 MWt and 1000 deg C outlet temperature is being developed by Japan Atomic Energy Research Institute and others since 1969, and the development of direct iron-making technology utilizing high temperature reducing gas was started in 1973 as the large project of Ministry of Internalional Trade and Industry. Kawasaki Heavy Industries, Ltd., Has taken part in these development projects, and has developed many softwares for nuclear heat design, system design and safety design of nuclear reactor system and heat utilization system. In hardwares also, efforts have been exerted to develop the technologies of design and manufacture of high temperature machinery and equipments. The high temperature, high pressure, large helium gas loop is under construction in the technical research institute of the company, and it is expected to be completed in December, 1979. The tests planned are that of proving the dynamic performances of the loop and its machinery and equipments and the verification of analysis codes. The loop is composed of the main circulation system, the objects of testing, the helium gas purifying system, the helium supplying and evacuating system, instruments and others. (Kako, I.)

  2. 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)

  3. Design and application for a high-temperature nuclear heat source

    International Nuclear Information System (INIS)

    Quade, R.N.

    1980-01-01

    Recent actions by OPEC have sharply increased interest in the United States in synfuels, with coal being the logical choice for the carbon source. Two coal liquefaction processes, direct and indirect, have been examined. Each can produce about 50% more output when coupled to an HTGR for process heat. The nuclear reactor designed for process heat has a power output of 842MW(t), a core outlet temperature of 950 0 C (1742 0 F), and an intermediate helium loop to separate the heat source from the process heat exchangers. Steam-methane reforming is the reference process. As part of the development of a nuclear process heat system, a computer code, Process Heat Reactor Evaluation and Design, is being developed. This code models both the reactor plant and a steam reforming plant. When complete, the program will have the capability to calculate an overall mass and heat balance, size the plant components, and estimate the plant cost for a wide variety of independent variables. (author)

  4. Neutronics and Thermal Hydraulics Analysis of a Conceptual Ultra-High Temperature MHD Cermet Fuel Core for Nuclear Electric Propulsion

    Directory of Open Access Journals (Sweden)

    Jian Song

    2018-04-01

    Full Text Available Nuclear electric propulsion (NEP offers unique advantages for the interplanetary exploration. The extremely high conversion efficiency of magnetohydrodynamics (MHD conversion nuclear reactor makes it a highly potential space power source in the future, especially for NEP systems. Research on ultra-high temperature reactor suitable for MHD power conversion is performed in this paper. Cermet is chosen as the reactor fuel after a detailed comparison with the (U,ZrC graphite-based fuel and mixed carbide fuel. A reactor design is carried out as well as the analysis of the reactor physics and thermal-hydraulics. The specific design involves fuel element, reactor core, and radiation shield. Two coolant channel configurations of fuel elements are considered and both of them can meet the demands. The 91 channel configuration is chosen due to its greater heat transfer performance. Besides, preliminary calculation of nuclear criticality safety during launch crash accident is also presented. The calculation results show that the current design can meet the safety requirements well.

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. High temperature material characterization and advanced materials development

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Kim, D. H.; Kim, S. H. and others

    2005-03-01

    The study is to characterize the structural materials under the high temperature, one of the most significant environmental factors in nuclear systems. And advanced materials are developed for high temperature and/or low activation in neutron irradiation. Tensile, fatigue and creep properties have been carried out at high temperature to evaluate the mechanical degradation. Irradiation tests were performed using the HANARO. The optimum chemical composition and heat treatment condition were determined for nuclear grade 316NG stainless steel. Nitrogen, aluminum, and tungsten were added for increasing the creep rupture strength of FMS steel. The new heat treatment method was developed to form more stable precipitates. By applying the novel whiskering process, high density SiC/SiC composites with relative density above 90% could be obtained even in a shorter processing time than the conventional CVI process. Material integrated databases are established using data sheets. The databases of 6 kinds of material properties are accessible through the home page of KAERI material division

  10. 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)

  11. 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

  12. High temperature corrosion in the thermochemical hydrogen production from nuclear heat

    International Nuclear Information System (INIS)

    Coen-Porisini, F.; Imarisio, G.

    1976-01-01

    In the production of hydrogen by water decomposition utilizing nuclear heat, a multistep process has to be employed. Water and the intermediate chemical products reach in chemical cycles giving hydrogen and oxygen with regeneration of the primary products used. Three cycles are examined, characterized by the presence of halide compounds and particularly hydracids at temperatures up to 800 0 C. Corrosion tests were carried out in hydrobromic acid, hydrochloric acid, ferric chloride solutions, and hydriodic acid

  13. 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

  14. 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.

  15. 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)

  16. Fiber Bragg Gratings for High-Temperature Thermal Characterization

    International Nuclear Information System (INIS)

    Stinson-Bagby, Kelly L.; Fielder, Robert S.

    2004-01-01

    Fiber Bragg grating (FBG) sensors were used as a characterization tool to study the SAFE-100 thermal simulator at the Nasa Marshal Space Flight Center. The motivation for this work was to support Nasa space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. Distributed high temperature measurements, up to 1150 deg. C, were made with FBG temperature sensors. Additionally, FBG strain measurements were taken at elevated temperatures to provide a strain profile of the core during operation. This paper will discuss the contribution of these measurements to meet the goals of Nasa Marshall Space Flight Center's Propulsion Research Center. (authors)

  17. 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)

  18. Extreme Temperature Radiation Tolerant Instrumentation for Nuclear Thermal Propulsion Engines, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of this proposal is to develop and commercialize a high reliability, high temperature smart neutron flux sensor for NASA Nuclear Thermal Propulsion...

  19. Energetic M1 transitions as a probe of nuclear collectivity at high temperatures

    International Nuclear Information System (INIS)

    Baktash, C.

    1987-01-01

    At ORNL, we have recently utilized the Spin Spectrometer setup to investigate the differential effects of increasing spin and excitation energy on nuclear shape and collectivity in 158 Yb. Along the yrast line of this and other N = 88 nuclei, weakly prolate shapes gradually give way to triaxial, and then finally to non-collective oblate shapes as the spin approaches 40 h-bar. However, above the yrast line, large deformation and collectivity once again sets in. This is evidenced by the emergence of a broad quadrupole structure (E/sub γ/ ≅ 1.2 MeV) in the continuum gamma-ray spectra that grows with increasing temperature. The short (sub ps) lifetimes of these transitions attest to the collective nature of these structures. The emergence and growth of the quadrupole structure at high excitation energies is closely correlated with the appearance of energetic (E/sub γ/ ≅ 2.5 MeV), fast M1 transitions which form another broad structure in the continuum spectra. From the centroid of the M1 bump, a quadrupole deformation parameter of 0.35 is inferred. Because of this sensitivity, these energetic M1 transitions provide a unique probe of nuclear shape in the excitation energy range of ≅ 3 to 10 MeV. 6 refs., 2 figs

  20. Basic thermo-fluid dynamic problems in high temperature heat exchangers

    International Nuclear Information System (INIS)

    McEligot, D.M.

    1986-01-01

    The authors consider high temperature heat exchangers to be ones where the heat transfer coefficients cannot be predicted confidently by classical analyses for pure forced convection with constant fluid properties. Alternatively, one could consider heat exchangers operating above some arbitrary temperature, say 1000F or 600C perhaps, to be at high temperature conditions. In that case, most common working fluids will be superheated vapors or gases. While some liquid metal heat exchangers are designed to operate in this range, the heat transfer coefficients of liquid metals are usually sufficiently high that the dominant thermal resistance would be due to the second fluid. This paper concentrates on convective heat transfer with gases. Typical applications include modular gas cooled nuclear reactors, proposed nuclear propulsion systems and space power plants, and superheaters in Rankine steam cycles

  1. 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

  2. Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

    International Nuclear Information System (INIS)

    Djati Hoesen Salimy

    2010-01-01

    The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

  3. 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.

  4. High Temperature Materials Interim Data Qualification Report FY 2011

    International Nuclear Information System (INIS)

    Lybeck, Nancy

    2011-01-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim fiscal year (FY) 2011 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under the Nuclear Quality Assurance (NQA)-1 guidelines and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from seven test series within the High Temperature Materials data stream have been entered into the NDMAS vault, including tensile tests, creep tests, and cyclic tests. Of the 5,603,682 records currently in the vault, 4,480,444 have been capture passed, and capture testing is in process for the remaining 1,123,238.

  5. Low-temperature nuclear orientation

    International Nuclear Information System (INIS)

    Stone, N.J.; Postma, H.

    1986-01-01

    This book comprehensively surveys the many aspects of the low temperature nuclear orientation method. The angular distribution of radioactive emissions from nuclei oriented by hyperfine interactions in solids, is treated experimentally and theoretically. A general introductory chapter is followed by formal development of the theory of the orientation process and the anisotropic emission of decay products from oriented nuclei, applied to radioactive decay and to reactions. Five chapters on applications to nuclear physics cover experimental studies of alpha, beta and gamma emission, nuclear moment measurement and level structure information. Nuclear orientation studies of parity non-conservation and time reversal asymmetry are fully described. Seven chapters cover aspects of hyperfine interactions, magnetic and electric, in metals, alloys and insulating crystals, including ordered systems. Relaxation phenomena and the combined technique of NMR detection using oriented nuclei are treated at length. Chapters on the major recent development of on-line facilities, giving access to short lived nuclei far from stability, on the use of nuclear orientation for thermometry below 1 Kelvin and on technical aspects of the method complete the main text. Extensive appendices, table of relevant parameters and over 1000 references are included to assist the design of future experiments. (Auth.)

  6. 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

  7. 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

  8. High temperature corrosion of metals

    International Nuclear Information System (INIS)

    Quadakkers, W.J.; Schuster, H.; Ennis, P.J.

    1988-08-01

    This paper covers three main topics: 1. high temperature oxidation of metals and alloys, 2. corrosion in sulfur containing environments and 3. structural changes caused by corrosion. The following 21 subjects are discussed: Influence of implanted yttrium and lanthanum on the oxidation behaviour of beta-NiA1; influence of reactive elements on the adherence and protective properties of alumina scales; problems related to the application of very fine markers in studying the mechanism of thin scale formation; oxidation behaviour of chromia forming Co-Cr-Al alloys with or without reactive element additions; growth and properties of chromia-scales on high-temperature alloys; quantification of the depletion zone in high temperature alloys after oxidation in process gas; effects of HC1 and of N2 in the oxidation of Fe-20Cr; investigation under nuclear safety aspects of Zircaloy-4 oxidation kinetics at high temperatures in air; on the sulfide corrosion of metallic materials; high temperature sulfide corrosion of Mn, Nb and Nb-Si alloys; corrosion behaviour or NiCrAl-based alloys in air and air-SO2 gas mixtures; sulfidation of cobalt at high temperatures; preoxidation for sulfidation protection; fireside corrosion and application of additives in electric utility boilers; transport properties of scales with complex defect structures; observations of whiskers and pyramids during high temperature corrosion of iron in SO2; corrosion and creep of alloy 800H under simulated coal gasification conditions; microstructural changes of HK 40 cast alloy caused by exploitation in tubes in steam reformer installation; microstructural changes during exposure in corrosive environments and their effect on mechanical properties; coatings against carburization; mathematical modeling of carbon diffusion and carbide precipitation in Ni-Cr-based alloys. (MM)

  9. 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

  10. 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.

  11. 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

  12. Losses of radionuclides related to high temperature ashing

    International Nuclear Information System (INIS)

    Carlsson, G.

    1985-01-01

    When measuring radionuclides in samples from the environment of nuclear power plants, a sample preparation step, such as high temperature ashing is often necessary. Althoug much used, this method is subject to controversy because of the risk of losses of several elements. A study, including the ashing of synthetically prepared samples has been undertaken. Controlled and moderate temperature rise rate and a final temperature not exceeding 550 deg. C has been shown vital for recovery

  13. 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

  14. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    International Nuclear Information System (INIS)

    Powell, J.; Reich, M.; Barletta, R.

    1996-01-01

    A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

  15. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Ballinger, R. [Massachusetts Institute of Technology (MIT); Majumdar, S. [Argonne National Laboratory (ANL); Weaver, K. D. [Idaho National Laboratory (INL)

    2008-03-01

    The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures

  16. Optimized Flow Sheet for a Reference Commercial-Scale Nuclear-Driven High-Temperature Electrolysis Hydrogen Production Plant

    International Nuclear Information System (INIS)

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

    2007-01-01

    This report presents results from the development and optimization of a reference commercial scale high-temperature electrolysis (HTE) plant for hydrogen production. The reference plant design is driven by a high-temperature helium-cooled 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 consists of 4.176 - 10 6 cells with a per-cell active area of 225 cm2. A nominal cell area-specific resistance, ASR, value of 0.4 Ohm-cm2 with a current density of 0.25 A/cm2 was used, and isothermal boundary conditions were assumed. 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 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 overall system thermal-to-hydrogen production efficiency (based on the low heating value of the produced hydrogen) is 49.07% at a hydrogen production rate of 2.45 kg/s with the high-temperature helium-cooled reactor concept. The information presented in this report is intended to establish an optimized design for the reference nuclear-driven HTE hydrogen production plant so that parameters can be compared with other hydrogen production methods and power cycles to evaluate relative performance characteristics and plant economics

  17. 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)

  18. Effects of porosity and temperature on oxidation behavior in air of selected nuclear graphites

    International Nuclear Information System (INIS)

    Chen Dongyue; Li Zhengcao; Miao Wei; Zhang Zhengjun

    2012-01-01

    Nuclear graphite endures gas oxidation in High Temperature Gas-cooled Reactor (HTGR), which may threaten the safety of reactor. To study the oxidation behavior of nuclear graphite, weight loss curve is usually measured through Thermo Gravimetric Analysis (TGA) method. In this work, three brands of nuclear graphite for HTGR (i.e., HSM-SC, IG-11, and NBG-18) are oxidized under 873 and 1073 K in open air, and their weight loss curves are obtained. The acceleration of oxidizing rate is observed for both HSM-SC and IG-11, and is attributed to the large porosity increase during oxidation process. For HSM-SC, the porosity increase comes from preferential binder oxidation, and thus its binder quality shall be improved to obtain better oxidation resistance. Temperature effects on oxidation for HSM-SC are also studied, which shows that oxidizing gas tends to be exhausted at graphite surface at high temperature instead of penetrate into the interior of bulk. (author)

  19. 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.)

  20. 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.)

  1. 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.)

  2. Low-temperature nuclear heat applications: Nuclear power plants for district heating

    International Nuclear Information System (INIS)

    1987-08-01

    The IAEA reflected the needs of its Member States for the exchange of information in the field of nuclear heat application already in the late 1970s. In the early 1980s, some Member States showed their interest in the use of heat from electricity producing nuclear power plants and in the development of nuclear heating plants. Accordingly, a technical committee meeting with a workshop was organized in 1983 to review the status of nuclear heat application which confirmed both the progress made in this field and the renewed interest of Member States in an active exchange of information about this subject. In 1985 an Advisory Group summarized the Potential of Low-Temperature Nuclear Heat Application; the relevant Technical Document reviewing the situation in the IAEA's Member States was issued in 1986 (IAEA-TECDOC-397). Programme plans were made for 1986-88 and the IAEA was asked to promote the exchange of information, with specific emphasis on the design criteria, operating experience, safety requirements and specifications for heat-only reactors, co-generation plants and power plants adapted for heat application. Because of a growing interest of the IAEA's Member States about nuclear heat employment in the district heating domaine, an Advisory Group meeting was organized by the IAEA on ''Low-Temperature Nuclear Heat Application: Nuclear Power Plants for District Heating'' in Prague, Czechoslovakia in June 1986. The information gained up to 1986 and discussed during this meeting is embodied in the present Technical Document. 22 figs, 11 tabs

  3. Comparison of microstructural and mechanical properties of joints developed by high temperature brazing, GTAW and laser welding methods on AISI 316 L stainless steel for specific applications in nuclear components

    International Nuclear Information System (INIS)

    Venkatesu, Sadu; Saxena, Rajesh; Ravi Kumar, R.; Chaurasia, P.K; Murugan, S.; Venugopal, S.

    2016-01-01

    Fabrication of instrumented irradiation capsule for evaluating the irradiation performance of fuel and structural materials in a nuclear reactor requires development of thin wall joints capable of withstanding high temperature and/or internal pressure. Thin wall joints for high temperature (∼550℃) applications can be made by laser beam welding (LBW), gas tungsten Arc welding (GTAW) and High Temperature Brazing (HLT) method

  4. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    Energy Technology Data Exchange (ETDEWEB)

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from

  5. Noninvasive ultrasonic measurements of temperature distribution and heat fluxes in nuclear systems

    International Nuclear Information System (INIS)

    Jia, Yunlu; Skliar, Mikhail

    2015-01-01

    Measurements of temperature and heat fluxes through structural materials are important in many nuclear systems. One such example is dry storage casks (DSC) that are built to store highly radioactive materials, such as spent nuclear reactor fuel. The temperature inside casks must be maintained within allowable limits of the fuel assemblies and the DSC components because many degradation mechanisms are thermally controlled. In order to obtain direct, real-time measurements of temperature distribution without insertion of sensing elements into harsh environment of storage casks, we are developing noninvasive ultrasound (US) methods for measuring spatial distribution of temperature inside solid materials, such as concrete overpacks, steel casings, thimbles, and rods. The measured temperature distribution can then be used to obtain heat fluxes that provide calorimetric characterisation of the fuel decay, fuel distribution inside the cask, its integrity, and accounting of nuclear materials. The physical basis of the proposed approach is the temperature dependence of the speed of sound in solids. By measuring the time it takes an ultrasound signal to travel a known distance between a transducer and a receiver, the indication about the temperature distribution along the path of the ultrasound propagation may be obtained. However, when temperature along the path of US propagation is non-uniform, the overall time of flight of an ultrasound signal depends on the temperature distribution in a complex and unknown way. To overcome this difficulty, the central idea of our method is to create an US propagation path inside material of interest which incorporates partial ultrasound reflectors (back scatterers) at known locations and use the train of created multiple echoes to estimate the temperature distribution. In this paper, we discuss experimental validation of this approach, the achievable accuracy and spatial resolution of the measured temperature profile, and stress the

  6. HTGR fuel behavior at very high temperature

    International Nuclear Information System (INIS)

    Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

    1986-03-01

    Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

  7. 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.)

  8. 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.

  9. Evaluation of corrosion inhibitors for high temperature decontamination applications

    International Nuclear Information System (INIS)

    Sathyaseelan, V.S.; Rufus, A.L.; Velmurugan, S.

    2015-01-01

    Normally, chemical decontamination of coolant systems of nuclear power reactors is carried out at temperatures less than 90 °C. At these temperatures, though magnetite dissolves effectively, the rate of dissolution of chromium and nickel containing oxides formed over stainless steel and other non-carbon steel coolant system surfaces is not that appreciable. A high temperature dissolution process using 5 mM NTA at 160 °C developed earlier by us was very effective in dissolving the oxides such as ferrites and chromites. However, the corrosion of structural materials such as carbon steel (CS) and stainless steel (SS) also increased beyond the acceptable limits at elevated temperatures. Hence, the control of base metal corrosion during the high temperature decontamination process is very important. In view of this, it was felt essential to investigate and develop a suitable inhibitor to reduce the corrosion that can take place on coolant structural material surfaces during the high temperature decontamination applications with weak organic acids. Three commercial inhibitors viz., Philmplus 5K655, Prosel PC 2116 and Ferroqest were evaluated at ambient and at 160 °C temperature in NTA formulation. Preliminary evaluation of these corrosion inhibitors carried out using electrochemical techniques showed maximum corrosion inhibition efficiency for Philmplus. Hence, it was used for high temperature applications. A concentration of 500 ppm was found to be optimum at 160 °C and at this concentration it showed an inhibition efficiency of 62% for CS. High temperature dissolution of oxides such as Fe 3 O 4 and NiFe 2 O 4 , which are relevant to nuclear reactors, was also carried out and the rate of dissolution observed was less in the presence of Philmplus. Studies were also carried out to evaluate hydrazine as a corrosion inhibitor for high temperature applications. The results revealed that for CS inhibition efficiency of hydrazine is comparable to that of Philmplus, while

  10. Study on temperature field airborne remote sensing survey along shore nuclear power station in different tide status

    International Nuclear Information System (INIS)

    Liang Chunli; Li Mingsong

    2010-01-01

    Nuclear Power Station needs to let large quantity of cooling water to the near sea area when it is running. Whether the cooling water has effect to surrounding environment and the running of Nuclear Power Station needs further research. Temperature Drainage Mathematic Model and Physical Analogue Model need to acquire the distribution characteristic of near Station sea surface temperature field in different seasons and different tide status. Airborne Remote Sending Technique has a advantage in gaining high resolution sea surface temperature in different tide status, and any other manual method with discrete point survey can not reach it. After a successful implementation of airborne remote sensing survey to gain the near-shore temperature drainage information in Qinshan Nuclear Power Station, it provides the reference methods and ideas for temperature drainage remote sensing survey of Nuclear Power Station. (authors)

  11. Temperature Stabilized Characterization of High Voltage Power Supplies

    CERN Document Server

    Krarup, Ole

    2017-01-01

    High precision measurements of the masses of nuclear ions in the ISOLTRAP experiment relies on an MR-ToF. A major source of noise and drift is the instability of the high voltage power supplies employed. Electrical noise and temperature changes can broaden peaks in time-of-flight spectra and shift the position of peaks between runs. In this report we investigate how the noise and drift of high-voltage power supplies can be characterized. Results indicate that analog power supplies generally have better relative stability than digitally controlled ones, and that the high temperature coefficients of all power supplies merit efforts to stabilize them.

  12. 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

  13. Heat exchangers for high-temperature thermodynamic cycles

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    The special requirements of heat exchangers for high temperature thermodynamic cycles are outlined and discussed with particular emphasis on cost and thermal stress problems. Typical approaches that have been taken to a comprehensive solution intended to meet all of the many boundary conditions are then considered by examining seven typical designs including liquid-to-liquid heat exchangers for nuclear plants, a heater for a closed cycle gas turbine coupled to a fluidized bed coal combustion chamber, steam generators for nuclear plants, a fossil fuel-fired potassium boiler, and a potassium condenser-steam generator. (auth)

  14. 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)

  15. Dynamic Nuclear Polarization at low temperature and high magnetic eld for biomedical applications in Magnetic Resonance Spectroscopic Imaging

    International Nuclear Information System (INIS)

    Goutailler, Florent

    2011-01-01

    The aim of this thesis work was to design, build and optimize a large volume multi-samples DNP (Dynamic Nuclear Polarization) polarizer dedicated to Magnetic Resonance Spectroscopic Imaging applications. The experimental system is made up of a high magnetic field magnet (3,35 T) in which takes place a cryogenic system with a pumped bath of liquid helium ("4He) allowing temperatures lower than 1,2 K. A set of inserts is used for the different steps of DNP: irradiation of the sample by a microwave field (f=94 GHz and P=50 mW), polarization measurement by Nuclear Magnetic Resonance... With this system, up to three samples of 1 mL volume can be polarized to a rate of few per-cents. The system has a long autonomy of four hours, so it can be used for polarizing molecules with a long time constant of polarization. Finally, the possibility to get quasi-simultaneously, after dissolution, several samples with a high rate of polarization opens the way of new applications in biomedical imaging. (author) [fr

  16. Divertor, thermonuclear device and method of neutralizing high temperature plasma

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

    The thermonuclear device comprises a thermonuclear reactor for taking place fusion reactions to emit fusion plasmas, and a divertor made of a hydrogen occluding material, and the divertor is disposed at a position being in contact with the fusion plasmas after nuclear fusion reaction. The divertor is heated by fusion plasmas after nuclear fusion reaction, and hydrogen is released from the hydrogen occluding material as a constituent material. A gas blanket is formed by the released hydrogen to cool and neutralize the supplied high temperature nuclear fusion plasmas. This prevents the high temperature plasmas from hitting against the divertor, elimination of the divertor by melting and evaporation, and solve a problem of processing a divertor activated by neutrons. In addition, it is possible to utilize hydrogen isotopes of fuels effectively and remove unnecessary helium. Inflow of impurities from out of the system can also be prevented. (N.H.)

  17. The temperature pulse propagation in interacting nuclear slabs

    International Nuclear Information System (INIS)

    Kozlowski, M.; Zurich Univ.

    1989-01-01

    Following the method developed by R.J. Swenson a non-Fourier equation for heat transfer in nuclear matter is obtained. The velocity of heat propagation is calculated and the value υ s = υ F /√3 is obtained. For two interacting nuclear slabs the temperature as a function of time is calculated. For a nucleon mean free path λ ≅ 3 fm the temperature saturation time is calculated and the value ≅ 160 fm/c is obtained. (orig.)

  18. Nuclear order in silver at pico-Kelvin temperature

    DEFF Research Database (Denmark)

    Siemensmeyer, K.; Clausen, K.N.; Lefmann, K.

    1997-01-01

    Nuclear order in silver is observed by neutron diffraction at pico-Kelvin temperatures. The structure is a type-I antiferromagnet with critical field of 100 mu T. The entropy-field phase diagram was determined using the spin-dependent absorption.......Nuclear order in silver is observed by neutron diffraction at pico-Kelvin temperatures. The structure is a type-I antiferromagnet with critical field of 100 mu T. The entropy-field phase diagram was determined using the spin-dependent absorption....

  19. 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)

  20. Effect of seawater and high-temperature history on swelling characteristics of bentonite

    International Nuclear Information System (INIS)

    Tanaka, Yukihisa; Nakamura, Kunihiko

    2005-01-01

    In the case of construction of repository for nuclear waste near the coastal area, the effect of seawater on swelling characteristics of bentonite as an engineering as an engineering barrier should be considered. Effects of high-temperature history on swelling characteristics of bentonite should also be considered because nuclear waste generates heat. Thus, in this study, swelling characteristics of bentonite on the conditions of high temperature history and seawater are investigated. The results of this study imply that : (1) Swelling strain of sodium bentonite or transformed sodium bentonite decrease as the salinity of water increases, whereas those of calcium bentonite are not affected by salinity of the water. (2) Quantitative evaluation method for swelling strain and swelling pressure of several kinds of bentonites under brine is proposed. (3) Using distilled water, swelling strain and swelling pressure of sodium bentonite with high-temperature history is less than those without high-temperature history. (author)

  1. Long duration performance of high temperature irradiation resistant thermocouples

    International Nuclear Information System (INIS)

    Rempe, J.; Knudson, D.; Condie, K.; Cole, J.; Wilkins, S.C.

    2007-01-01

    Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C degrees. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL's recommended thermocouple design, a series of high temperature (from 1200 to 1800 C) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 C that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests. (authors)

  2. Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

    International Nuclear Information System (INIS)

    Uchida, Shunsuke

    2008-01-01

    The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., increased shutdown radiation, generation of defects in materials of major components and fuel claddings, and increased volume of radwaste sources. Corrosion behavior is greatly affected by water quality and differs according to the water quality values and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of the key issues that determine corrosion-related problems, but it is not the only issue. Most corrosion-related phenomena, e.g., flow accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., the electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, the ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. (orig.)

  3. High temperature measurement by noise thermometry

    International Nuclear Information System (INIS)

    Decreton, M.C.

    1982-06-01

    Noise thermometry has received a lot of attention for measurements of temperatures in the high range around 1000-2000 deg. K. For these measurements, laboratory type experiments have been mostly performed. These have shown the interest of the technique when long term stability, high precision and insensibility to external conditions are concerned. This is particularly true for measurements in nuclear reactors where important drifts due to irradiation effects are experienced with other measurement techniques, as thermocouple for instance. Industrial noise thermometer experiments have not been performed extensively up to now. The subject of the present study is the development of a 1800 deg. K noise thermometer for nuclear applications. The measurement method is based on a generalized noise power approach. The rms noise voltage (Vsub(s)) and noise current (Isub(s)) are successively measured on the resistive sensor. The same quantities are also measured on a dummy short circuited probe (Vsub(d) and Isub(d)). The temperature is then deduced from these measured values by the following formula: cTsub(s) = (Vsub(s) 2 - Vsub(d) 2 )(Vsub(s)/Isub(s) - Vsub(d)/Isub(d)) - 1 , where c is a constant and Tsub(s) the absolute temperature of the sensor. This approach has the particular advantage of greatly reducing the sensibility to environmental perturbations on the leads and to the influence of amplifier noise sources. It also eliminates the necessity of resistance measurement and keeps the electronic circuits as simple as possible

  4. Microstructural characteristics of high-temperature oxidation in nickel-base superalloy

    International Nuclear Information System (INIS)

    Khalid, F.A.

    1997-01-01

    Superalloys are used for aerospace and nuclear applications where they can withstand high-temperature and severe oxidizing conditions. High-temperature oxidation behavior of a nickel-base superalloy is examined using optical and scanning electron microscopical techniques. The morphology of the oxide layers developed is examined, and EDX microanalysis reveals diffusion of the elements across the oxide-metal interface. Evidence of internal oxidation is presented, and the role of structural defects is considered. The morphology of the oxide-metal interface formed in the specimens exposed in steam and air is examined to elucidate the mechanism of high-temperature oxidation

  5. 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)

  6. 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

  7. 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)

  8. High-temperature flaw assessment procedure

    International Nuclear Information System (INIS)

    Ruggles, M.B.; Takahashi, Y.; Ainsworth, R.A.

    1991-08-01

    Described is the background work performed jointly by the Electric Power Research Institute in the United States, the Central Research Institute of Electric Power Industry in Japan and Nuclear Electric plc in the United Kingdom with the purpose of developing a high-temperature flaw assessment procedure for reactor components. Existing creep-fatigue crack-growth models are reviewed, and the most promising methods are identified. Sources of material data are outlined, and results of the fundamental deformation and crack-growth tests are discussed. Results of subcritical crack-growth exploratory tests, creep-fatigue crack-growth tests under repeated thermal transient conditions, and exploratory failure tests are presented and contrasted with the analytical modeling. Crack-growth assessment methods are presented and applied to a typical liquid-metal reactor component. The research activities presented herein served as a foundation for the Flaw Assessment Guide for High-Temperature Reactor Components Subjected to Creep-Fatigue Loading published separately. 30 refs., 108 figs., 13 tabs

  9. Application of Self-Propagating High Temperature Synthesis to the Fabrication of Actinide Bearing Nitride and Other Ceramic Nuclear Fuels

    International Nuclear Information System (INIS)

    Moore, John J.; Reigel, Marissa M.; Donohoue, Collin D.

    2009-01-01

    The project uses an exothermic combustion synthesis reaction, termed self-propagating high-temperature synthesis (SHS), to produce high quality, reproducible nitride fuels and other ceramic type nuclear fuels (cercers and cermets, etc.) in conjunction with the fabrication of transmutation fuels. The major research objective of the project is determining the fundamental SHS processing parameters by first using manganese as a surrogate for americium to produce dense Zr-Mn-N ceramic compounds. These fundamental principles will then be transferred to the production of dense Zr-Am-N ceramic materials. A further research objective in the research program is generating fundamental SHS processing data to the synthesis of (i) Pu-Am-Zr-N and (ii) U-Pu-Am-N ceramic fuels. In this case, Ce will be used as the surrogate for Pu, Mn as the surrogate for Am, and depleted uranium as the surrogate for U. Once sufficient fundamental data has been determined for these surrogate systems, the information will be transferred to Idaho National Laboratory (INL) for synthesis of Zr-Am-N, Pu-Am-Zr-N and U-Pu-Am-N ceramic fuels. The high vapor pressures of americium (Am) and americium nitride (AmN) are cause for concern in producing nitride ceramic nuclear fuel that contains Am. Along with the problem of Am retention during the sintering phases of current processing methods, are additional concerns of producing a consistent product of desirable homogeneity, density and porosity. Similar difficulties have been experienced during the laboratory scale process development stage of producing metal alloys containing Am wherein compact powder sintering methods had to be abandoned. Therefore, there is an urgent need to develop a low-temperature or low-heat fuel fabrication process for the synthesis of Am-containing ceramic fuels. Self-propagating high temperature synthesis (SHS), also called combustion synthesis, offers such an alternative process for the synthesis of Am nitride fuels. Although SHS

  10. Temperature Profile in Fuel and Tie-Tubes for Nuclear Thermal Propulsion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vishal Patel

    2015-02-01

    A finite element method to calculate temperature profiles in heterogeneous geometries of tie-tube moderated LEU nuclear thermal propulsion systems and HEU designs with tie-tubes is developed and implemented in MATLAB. This new method is compared to previous methods to demonstrate shortcomings in those methods. Typical methods to analyze peak fuel centerline temperature in hexagonal geometries rely on spatial homogenization to derive an analytical expression. These methods are not applicable to cores with tie-tube elements because conduction to tie-tubes cannot be accurately modeled with the homogenized models. The fuel centerline temperature directly impacts safety and performance so it must be predicted carefully. The temperature profile in tie-tubes is also important when high temperatures are expected in the fuel because conduction to the tie-tubes may cause melting in tie-tubes, which may set maximum allowable performance. Estimations of maximum tie-tube temperature can be found from equivalent tube methods, however this method tends to be approximate and overly conservative. A finite element model of heat conduction on a unit cell can model spatial dependence and non-linear conductivity for fuel and tie-tube systems allowing for higher design fidelity of Nuclear Thermal Propulsion.

  11. Study on elastic-plastic fracture toughness test in high temperature water

    International Nuclear Information System (INIS)

    Miura, Yasufumi

    2016-01-01

    Structural integrity of internal components in light water reactors is important for the safety of operation and service lifetime. Fracture toughness is important parameter for structural integrity assessment of nuclear power plant. In general, fracture toughness of materials which compose the components in light water reactor is obtained with fracture toughness tests in air although some components are subjected to high temperature water because of the difficulty of fracture toughness test in high temperature water. However, the effects of high temperature water and hydrogen on fracture behavior of the structural materials in nuclear power plant such as low alloy steel, cast austenitic stainless steel, and Ni base alloy are concerned recently. In this study, elastic-plastic fracture toughness test of low alloy steel in simulated BWR water environment was studied. Fracture toughness test in high temperature water with original clip gage and normalization data reduction technique was established. The difference of fracture toughness J_Q tested in air between using elastic unload compliance method and normalization data reduction technique was also discussed. As a result, obtained value with normalization data reduction technique tended to be higher than the value with elastic unload compliance. (author)

  12. 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.

  13. Temperature-dependent errors in nuclear lattice simulations

    International Nuclear Information System (INIS)

    Lee, Dean; Thomson, Richard

    2007-01-01

    We study the temperature dependence of discretization errors in nuclear lattice simulations. We find that for systems with strong attractive interactions the predominant error arises from the breaking of Galilean invariance. We propose a local 'well-tempered' lattice action which eliminates much of this error. The well-tempered action can be readily implemented in lattice simulations for nuclear systems as well as cold atomic Fermi systems

  14. The high temperature out-of-pile test of LVDT for elongation measurement of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Jo, M. S.; Joo, K. N.; Park, S. J.; Gang, Y. H.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the elongation measurement technique of the fuel pellet is being developed using LVDT(Linear Variable Differential Transformer). The well qualified out-of-pile test were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation instrumented capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for fuel pellet elongation was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C with increasing the elongation from 0 to 5 mm. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT and elongation detector at high temperature was introduced. It is known that the results will be used to predict accurately the elongation of fuel pellet during irradiation test.

  15. Application of zirconia membranes as high-temperature PH sensors

    International Nuclear Information System (INIS)

    Neidrach, L.W.

    1983-01-01

    The zirconia pH sensor behaves much like the classical glass electrode, but it extends the range of measurement to much higher temperatures - about 300 0 vs 120 0 C. It also has virtues over the glass electrode at lower temperatures because of the absence of an ''alkaline error.'' Like the glass electrode, it is insensitive to changes in the redox potential of the environment and, in turn, it exerts no influence on the environment. Such sensors have been finding application in the direct measurement of the pH of geothermal brines, of water in nuclear reactors, and in high-temperature corrosion studies. The sensors can also be used as ''pseudoreference'' electrodes for the measurement of redox and corrosion potentials in high-temperature media

  16. 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

  17. High and very high temperature reactor research for multipurpose energy applications

    International Nuclear Information System (INIS)

    Hittner, Dominique; Bogusch, Edgar; Fuetterer, Michael; Groot, Sander de; Ruer, Jacques

    2011-01-01

    Ten years ago, the European High Temperature Reactor (HTR) Technology Network (HTR-TN) launched a programme for developing HTR Technology, which expanded so far through 4 successive Euratom Framework Programmes. Many projects have been performed - in particular the RAPHAEL project in the 6th Euratom Framework Programme and presently ARCHER in the 7th - in line with the Network strategy that identified cogeneration of process heat and power as the main specific mission of HTR. HTR can indeed address the growing energy needs of industry presently fully relying on fossil fuel combustion with a CO 2 -lean generation technology, thanks to its high operating temperature and to its unique flexibility obtained from its large thermal inertia and its low power. Relying on the legacy of the former European leadership in HTR technology, this programme has addressed specific developments required for industrial process heat applications and for increasing HTR performances (higher temperatures and fuel burn-up). Decisive achievements have been obtained concerning fuel manufacturing and irradiation behaviour, key components and their materials, safety, computer code validation and specific HTR waste (fuel and graphite) management. Key experiments have been performed or are still ongoing: irradiation of graphite, fuel and vessel materials and the corresponding post-irradiation examinations, safety tests and isotopic analyses; thermal-hydraulic tests of an Intermediate Heat Exchanger mock-up in helium; air ingress experiments for a block type core, etc. Through Euratom participation in the Generation IV International Forum (GIF), these achievements contribute to international cooperation. HTR-TN strategy has been recently integrated by the 'Sustainable Nuclear Energy Technology Platform' (SNE-TP) as one of the 3 'pillars' of its global nuclear strategy. It is also in line with the orientations and the timing of the 'Strategic Energy Technology Plan (SET-Plan)' for the development

  18. How is it possible to measure a nuclear temperature

    International Nuclear Information System (INIS)

    Tamain, B.

    1989-01-01

    Several methods for the measurement of nuclear temperatures are summarized. The concepts of hot nuclei and temperature are defined. The nuclear equation of state is presented. The statistical theory of hot nuclei decay properties is analyzed. The obtention of the excitation energy from the recoil velocity measurement is considered in the case of complete and incomplete fusion. The measurements of temperature and excitation energy from the properties of decay products are reviewed. The study shows that no measurement method is perfect. Moreover, it is necessary to select events for which the degree of dissipation of the incident energy is estimated

  19. The materials programme for the high-temperature gas-cooled reactor in the Federal Republic of Germany: Status of the development of high-temperature materials, integrity concept, and design codes

    International Nuclear Information System (INIS)

    Nickel, H.; Bodmann, E.; Seehafer, H.J.

    1990-01-01

    During the last 15 years, the research and development of materials for high temperature gas-cooled reactor (HTGR) applications in the Federal Republic of Germany have been concentrated on the qualification of high-temperature structural alloys. Such materials are required for heat exchanger components of advanced HTGRs supplying nuclear process heat in the temperature range between 750 deg. and 950 deg. C. The suitability of the candidate alloys for service in the HTGR has been established, and continuing research is aimed at verification of the integrity of components over the envisaged service lifetimes. The special features of the HTGR which provide a high degree of safety are the use of ceramics for the core construction and the low power density of the core. The reactor integrity concept which has been developed is based on these two characteristics. Previously, technical guidelines and design codes for nuclear plants were tailored exclusively to light water reactor systems. An extensive research project was therefore initiated which led to the formulation of the basic principles on which a high temperature design code can be based. (author)

  20. High-temperature thermophysical properties of nuclear fuels

    International Nuclear Information System (INIS)

    Hyland, G.J.; Ralph, Jeffrey

    1985-01-01

    This is a summary of discussions held at the 9th European Thermophysical Properties Conference. The first group discussed the following: the question of Bredig transition in UO 2 and related oxides, the thermal conductivity of molten UO 2 , the status of first principle calculations of the free energies of defect formation in UO 2 . A second group of topics discussed were: oxygen potentials over mixed oxide systems, the current status of vapor pressure measurements over liquid UO 2 made by use of laser heating techniques and their interpretation and preliminary results on electric and dielectric properties of solid UO 2 at high frequencies and low temperatures. The main interest was in the thermal conductivity of molten UO 2 and much of the discussion time was devoted to this. (U.K.)

  1. Multiplexing milli-volt transmitter for operation in high ambient temperatures

    International Nuclear Information System (INIS)

    Phillips, G.J.

    1980-01-01

    A high integrity method of multiplexing up to two hundred and fifty millivolt level signals and transmitting the data to a remote measuring station via a 12 core flexible cable is described. The system was designed for operation in the normally hazardous and therefore inaccessible areas where high ambient temperatures are experienced. Additionally, because one potential application is in nuclear reactor systems, the design is tolerant to high levels of gamma background. The system's high reliability, high integrity and relatively small and conventional cable installation, makes it applicable to situations which depend upon temperature measurement for plant or personnel safety. (author)

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. High temperature x-ray micro-tomography

    Energy Technology Data Exchange (ETDEWEB)

    MacDowell, Alastair A., E-mail: aamacdowell@lbl.gov; Barnard, Harold; Parkinson, Dilworth Y.; Gludovatz, Bernd [Lawrence Berkeley National Lab., Berkeley, CA 94720 (United States); Haboub, Abdel [Lawrence Berkeley National Lab., Berkeley, CA 94720 (United States); current –Lincoln Univ., Jefferson City, Missouri, 65101 (United States); Larson, Natalie; Zok, Frank [University California Santa Barbara, Santa Barbara CA 93106 (United States); Panerai, Francesco; Mansour, Nagi N. [NASA Ames Research Centre, Moffett Field, CA, 94035 (United States); Bale, Hrishikesh [University California Berkeley, Berkeley, CA 94720 (United States); current - Carl Zeiss X-ray Microscopy, 4385 Hopyard Rd #100, Pleasanton, CA 94588 (United States); Acevedo, Claire [Lawrence Berkeley National Lab., Berkeley, CA 94720 (United States); University California San Francisco, San Francisco, CA 94143 (United States); Liu, Dong [University of Bristol, Bristol BS8 1TH (United Kingdom); Ritchie, Robert O. [Lawrence Berkeley National Lab., Berkeley, CA 94720 (United States); University California Berkeley, Berkeley, CA 94720 (United States)

    2016-07-27

    There is increasing demand for 3D micro-scale time-resolved imaging of samples in realistic - and in many cases extreme environments. The data is used to understand material response, validate and refine computational models which, in turn, can be used to reduce development time for new materials and processes. Here we present the results of high temperature experiments carried out at the x-ray micro-tomography beamline 8.3.2 at the Advanced Light Source. The themes involve material failure and processing at temperatures up to 1750°C. The experimental configurations required to achieve the requisite conditions for imaging are described, with examples of ceramic matrix composites, spacecraft ablative heat shields and nuclear reactor core Gilsocarbon graphite.

  7. Status of high-temperature heat-pipe technology

    International Nuclear Information System (INIS)

    Ranken, W.A.

    1982-01-01

    This paper discusses the application of heat pipes to nuclear reactor space power systems. Characteristics of the device that favor such an application are described and recent results of current technology development programs are presented. Research areas that will need to be addressed in demonstrating that adequate lifetimes can be achieved with evaporation/condensation cycles operating at high temperatures in a reactor environment are also discussed

  8. Radiation and temperature effects on electronic components investigated under the CSTI high capacity power project

    International Nuclear Information System (INIS)

    Schwarze, G.E.; Niedra, J.M.; Frasca, A.J.; Wieserman, W.R.

    1993-01-01

    The effects of nuclear radiation and high temperature environments must be fully known and understood for the electronic components and materials used in both the Power Conditioning and Control subsystem and the reactor Instrumentation and Control subsystem of future high capacity nuclear space power systems. This knowledge is required by the designer of these subsystems in order to develop highly reliable, long-life power systems for future NASA missions. A review and summary of the experimental results obtained for the electronic components and materials investigated under the power management element of the CSTI high capacity power project will be presented in this paper: (1) Neutron, gamma ray, and temperature effects on power semiconductor switches, (2) Temperature and frequency effects on soft magnetic materials; and (3) Temperature effects on rare-earth permanent magnets

  9. Radiation and temperature effects on electronic components investigated under the CSTI High Capacity Power Project

    International Nuclear Information System (INIS)

    Shwarze, G.E.; Wieserman, W.R.

    1994-01-01

    The effects of nuclear radiation and high temperature environments must be fully known and understood for the electronic components and materials used in both the Power Conditioning and Control subsystem and the reactor Instrumentation and Control subsystem of future high capacity nuclear space power systems. This knowledge is required by the designer of these subsystems in order to develop highly reliable, long-life power systems for future NASA missions. A review and summary of the experimental results obtained for the electronic components and materials investigated under the power management element of the CSTI high capacity power project will be presented in this paper: (1) Neutron, gamma ray, and temperature effects on power semiconductor switches, (2) Temperature and frequency effects on soft magnetic materials; and (3) Temperature effects on rare earth permanent magnets

  10. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    Science.gov (United States)

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. A multipurpose pollution-free high temperature heat supply system for 21st century service

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1996-01-01

    In the 21st century, increasing environmental concerns, together with decreasing fossil fuel resources, will result in a gradual transition in the power industry to the use of nuclear energy on a global scale. The demand for energy to meet growing populations and the needs of industry, transportation, and the heating market, will be based on the increasing use of electricity and hydrogen, these being produced, first by fission and later by fusion reactors. The realization of this scenario will be the deployment of a high temperature reactor (HTR), which together with a heat transport loop constitutes a nuclear heat source (NHS). The initial large-scale use of the NHS will likely be for nuclear process heat, namely the fossil-free production of hydrogen by thermochemical water splitting. The same NHS will also be used for the high efficiency generation of electricity using an indirect cycle helium gas turbine. An important stepping stone towards this goal will be the operation of a high temperature test reactor (HTTR) currently under construction in Japan. This will pave the way for introduction of the HTR for hydrogen production and electricity generation around the year 2020. This paper puts into perspective technological aspects of a futuristic, pollution free, high temperature nuclear heat source

  17. 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

  18. High-temperature thermal-chemical analysis of nuclear fuel channels

    Energy Technology Data Exchange (ETDEWEB)

    Nekhamkin, Y; Rosenband, V; Hasan, D; Elias, E; Wacholder, E; Gany, A [Technion-Israel Inst. of Tech., Haifa (Israel)

    1996-12-01

    In a severe accident situation, e.g., a postulated loss of coolant accident with a coincident loss of emergency core cooling (LOCA/LOECC), the core may become partially uncovered and steam may become the only coolant available. The thermodynamic conditions in the core, in this case, depend on ability of the steam to effectively remove the fuel decay heat and the heat generated by the exothermic steam/Zircaloy reaction., Therefore, it is important to understand the high-temperature behavior of an oxidizing fuel channel. The main objective of this work is to develop a methodology for calculating the clad temperature and rate of oxidation of a partially covered fuel pin. A criterion is derived to define the importance of the chemical reaction in the overall heat balance. The main parameters affecting the fuel thermal behavior are outlined (authors).

  19. Development and high temperature sealing performance study of double sealing flange

    International Nuclear Information System (INIS)

    Zhang Qinhua; Fu Yuan; Xie Leidong; Li Zhong; Tang Zhongfeng

    2013-01-01

    Background: Flange is an important component used in the connection between pipes or between pipe and equipment. It is widely used in mechanical industry, petrochemical industry and nuclear industry. Ordinary flange could be oxidized oi corroded under high temperature and strong corrosion condition, which leads to flange's failure. Purpose: The aim is to design and validate the double sealing flange which could be used in high temperature and strong corrosion condition stably and supplied as a reliable detachable connection for petrochemical industry and nuclear industry under high temperature and strong corrosion environment. Methods: New flange is designed using flexible graphite spiral wound gaskets and C style metal ring double sealing structure. The unique protection gas design ensures the graphite spiral wound gasket used in high temperature reliably. Test bed is built to verify the stability of the flange. Results: Pressure reducing of graphite spiral wound gasket is only 0.01 MPa at 750℃, 0.7 MPa for 43 h, there is not obvious oxidation phenomenon at the same time. The leakage rate of metal C sealing ring is 11.51×10 -4 atm cc/s, and the metal C ring sealing's performance meets the engineering requirements. Conclusion: The leakage rate of double sealing flange is almost zero under 0.7 MPa at 750℃, which meets the engineering requirements. It avoids oxidizing at high temperature and corroding under strong corrosion condition. (authors)

  20. The rectal temperature estimation method based on tympanic temperature for workers wearing protective clothing in nuclear facilities

    International Nuclear Information System (INIS)

    Takahashi, Naoki; Lee, Joo-Young; Wakabayashi, Hitoshi; Tochihara, Yutaka

    2012-01-01

    At nuclear facilities, workers wear impermeable protective clothing to prevent radioactive contamination during inspection and maintenance activities. The heat stroke risk of the workers wearing protective clothing gradually increases, because of retaining heat and humidity inside of protective clothing. Normally, the rectal temperature is used to manage the heat stroke risk. But the rectal temperature measurement is very difficult at the working place. We have already reported that the measurement of infrared tympanic temperature is more realistic than that of rectal temperature to manage the heat stroke risk. But tympanic temperature indicates high temperature compared to rectal temperature. So, the use of the tympanic temperature overestimates core temperature and decreases the work efficiency. Therefore, we attempted to make formulas to predict rectal temperature from measured tympanic temperature, and to use calculated rectal temperature for safer and more efficient management. The rectal temperature predicted with the formulas agreed with the actual measurement within the range of measurement error (±0.1degC). Combination of tympanic temperature measurement and heat rate evaluation enabled the safer management of the heat stroke risk with wearing protective clothing. (author)

  1. Container floor at high temperatures

    International Nuclear Information System (INIS)

    Reutler, H.; Klapperich, H.J.; Mueller-Frank, U.

    1978-01-01

    The invention describes a floor for container which is stressed at high, changing temperatures and is intended for use in gas-cooled nuclear reactors. Due to the downward cooling gas flow in these types of reactor, the reactor floor is subjected to considerable dimensional changes during switching on and off. In the heating stage, the whole graphite structure of the reactor core and floor expands. In order to avoid arising constraining forces, sufficiently large expansion spaces must be allowed for furthermore restoring forces must be present to close the gaps again in the cooling phase. These restoring forces must be permanently present to prevent loosening of the core cuits amongst one another and thus uncontrollable relative movement. Spring elements are not suitable due to fast fatigue as a result of high temperatures and radiation exposure. It is suggested to have the floor elements supported on rollers whose rolling planes are downwards inclined to a fixed point for support. The construction is described in detail by means of drawings. (GL) [de

  2. Design and installation of high-temperature ultrasonic measuring system and grinder for nuclear fuel containing trans-uranium elements

    International Nuclear Information System (INIS)

    Serizawa, Hiroyuki; Kikuchi, Hironobu; Iwai, Takashi; Arai, Yasuo; Kurosawa, Makoto; Mimura, Hideaki; Abe, Jiro

    2005-07-01

    A high-temperature ultrasonic measuring system had been designed and installed in a glovebox (711-DGB) to study a mechanical property of nuclear fuel containing trans-uranium (TRU) elements. A figuration apparatus for the cylinder-type sample preparation had also been modified and installed in an established glovebox (142-D). The system consists of an ultrasonic probe, a heating furnace, cooling water-circulating system, a cooling air compressor, vacuum system, gas supplying system and control system. An A/D converter board and an pulsar/receiver board for the measurement of wave velocity were installed in a personal computer. The apparatus was modified to install into the glovebox. Some safety functions were supplied to the control system. The shape and size of the sample was revised to minimize the amount of TRU elements for the use of the measurement. The maximum sample temperature is 1500degC. The performance of the installed apparatuses and the glovebox were confirmed through a series of tests. (author)

  3. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

    International Nuclear Information System (INIS)

    Huang Weiqing; Wu Ruixian; Zheng Yukuan

    2011-01-01

    In the dry storage of spent nuclear fuels,concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity,and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94 degree C for 90 days indicate that: (1) compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure,and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials,including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement. (authors)

  4. Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

    International Nuclear Information System (INIS)

    Uchida, Shunsuke

    2014-01-01

    The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., generating defects in materials of major components and fuel claddings, increasing shutdown radiation and increasing the volume of radwaste sources. Corrosion behaviors are much affected by water qualities and differ according to the values of water qualities and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of key issues that determine corrosion related problems but it is not the only issue. Most phenomena for corrosion related problems, e.g., flow-accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. In the paper, theoretical models based on electrochemistry to estimate wall thinning rate of carbon steel piping due to flow-accelerated corrosion and corrosive conditions determining IGSCC crack initiation and growth rate are introduced. (author)

  5. 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

  6. Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

    International Nuclear Information System (INIS)

    Sales, B.C.; Boatner, L.A.

    1984-01-01

    Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

  7. 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

  8. Potential of low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    1986-12-01

    At present, more than one third of the fossil fuel currently used is being consumed to produce space heating and to meet industrial needs in many countries of the world. Imported oil still represents a large portion of this fossil fuel and despite its present relatively low price future market evolutions with consequent upward cost revisions cannot be excluded. Thus the displacement of the fossil fuel by cheaper low-temperature heat produced in nuclear power plants is a matter which deserves careful consideration. Technico-economic studies in many countries have shown that the use of nuclear heat is fully competitive with most of fossil-fuelled plants, the higher investment costs being offset by lower production cost. Another point in favour of heat generation by nuclear source is its indisputable advantage in terms of benefits to the environment. The IAEA activity plans for 1985-86 concentrate on information exchange with specific emphasis on the design criteria, operating experience, safety requirements and specifications of heat-only reactors, co-generation plants and existing power plants backfitted for additional heat applications. The information gained up to 1985 was discussed during the Advisory Group Meeting on the Potential of Low-Temperature Nuclear Heat Applications held in the Federal Institute for Reactor Research, Wuerenlingen, Switzerland in September 1985 and, is included in the present Technical Document

  9. Practical reasons for investigating ion transport in high temperature insulating materials

    International Nuclear Information System (INIS)

    Sonder, E.

    1976-01-01

    Practical problems encountered in a number of advanced technology appliations, particularly those related to energy conversion, are discussed. Refractory ionic compounds which are abundant and of high melting point are listed, and technological problems are discussed in terms of specific materials problems. The argument is made that basic information concerning transport properties in refractory compounds is lacking to such an extent that it is difficult to design and assess advanced energy generation systems. Technology applications include: a) ceramic nuclear fuels for high temperature fission reactors, b) high temperature gas turbine blades, c) insulators in controlled thermonuclear reactors, and d) magnetohydrodynamic generators. Some of the difficulties inherent in making transport property measurements at high temperatures are also listed

  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. Pipe connection for high pressure and temperature loads

    International Nuclear Information System (INIS)

    Haferkamp, D.; Hodzic, A.; Paetz, E.; Stach, H.

    1976-01-01

    The patent proposes an inprovement of the clamping device for a pipe joint connecting pipelines which are subject to high pressure and temperature loads, e.g. in a nuclear power plant. This clamping device may be tightened and loosened by remote control. The proposed clamping ring consists of several segments connected with each other by hinge-type guide pins and fishplates. (UWI) [de

  12. Control system for high-temperature slagging incinerator plant

    International Nuclear Information System (INIS)

    Matsuzaki, Yuji

    1986-01-01

    Low-level radioactive wastes generated in the nuclear generating plants are increasing year by year and to dispose them safely constitutes a big problem for the society. A few years ago, as the means of reducing them to as little volume as possible by incinerating and fusing the wastes, a high-temperature slagging incinerating method was developed, and this method is highly assessed. JGC Corp. has introduced that system technology and in order to prove the capacity of disposal and salubrity of the plant, and have constructed a full-sized pilot plant, then obtained the operational record and performance as they had planned. This report introduces the general processing of the wastes from their incineration and fusion as well as process control technology characteristic to high-temperature slagging incinerator furnaces and sensor technology. (author)

  13. 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)

  14. High temperature mass spectrometry for thermodynamic study of radioactive materials

    International Nuclear Information System (INIS)

    Pattoret, Andre; Philippot, Joseph; Pesme, Olivier.

    1983-01-01

    Thermodynamic properties and evaporation kinetics are essential data to evaluate the nuclear fuel behaviour under accidental conditions. High temperature mass spectrometry appears as a valuable method to set up a such assessment. However, because of size, complexity and radioactivity of the irradiated samples, important improvements of the classical method are required. The device built in CEN/FAR to overcome these problems is described; performances and possible applications out of the nuclear safety field are presented [fr

  15. Status and prospects on development of yttrium-based high-temperature superconducting coated conductor

    International Nuclear Information System (INIS)

    Izumi, Teruo; Yanagi, Nagato

    2017-01-01

    Development of a large-sized large-current conductor using a high-temperature superconducting wire rod based on copper oxide has been started worldwide for the purpose of applying it as an option of a magnet for a nuclear fusion prototype reactor. There is yttrium-based thin film wire rod as a promising candidate. Japan is leading the development of this wire rod for many years, aiming to apply it to power equipment and the like. This paper explained the history of wire rod development, basic superconducting properties and manufacturing method, and latest achievements, and overviewed the feasibility of application to nuclear fusion reactor magnets. At present, the use of niobium-based low-temperature superconducting wire rod that is used in ITER is the basic idea. On the other hand, the development of wire rod using a copper oxide type high-temperature superconductor (HTS) has also been started. HTS wire rod is evaluated as suitable for application to nuclear fusion magnets due to its superior critical current characteristics and high mechanical rigidity up to high magnetic fields at high temperatures of yttrium. As current development progress, there are development of high-quality wire rod in the magnetic field and development of low AC loss wire rod. As future prospects, cost reduction due to mass production and improvement of yield, and investigation of low-resistance connection technology are being studied. The remaining future challenges of yttrium-based HTS are improvement of the anisotropy, influence on neutron irradiation, and problem of activation. (A.O.)

  16. Pressure pressure-balanced pH sensing system for high temperature and high pressure water

    International Nuclear Information System (INIS)

    Tachibana, Koji

    1995-01-01

    As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

  17. 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)

  18. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  19. Steam generators secondary side chemical cleaning at Point Lepreau using the Siemen's high temperature process

    International Nuclear Information System (INIS)

    Verma, K.; MacNeil, C.; Odar, S.

    1996-01-01

    The secondary sides of all four steam generators at the Point Lepreau Nuclear Generating Stations were cleaned during the 1995 annual outage run-down using the Siemens high temperature chemical cleaning process. Traditionally all secondary side chemical cleaning exercises in CANDU as well as the other nuclear power stations in North America have been conducted using a process developed in conjunction with the Electric Power Research Institute (EPRI). The Siemens high temperature process was applied for the first time in North America at the Point Lepreau Nuclear Generating Station (PLGS). The paper discusses experiences related to the pre and post award chemical cleaning activities, chemical cleaning application, post cleaning inspection results and waste handling activities. (author)

  20. Nuclear collective states at finite temperature

    International Nuclear Information System (INIS)

    Milian, A.; Barranco, M.; Mas, D.; Lombard, R.J.

    1987-04-01

    The Energy Density Method (EDM) has been used to study low-lying nuclear collective states as well as isoscalar giant resonances at finite temperature (T). Giant states have been studied by computing the corresponding strength function moments (sum rules) in the Random-Phase Approximation (RPA). For the description of the low lying states we have resorted to a variety of models from the rather sophisticated RPA method to liquid drop and schematic models. It has been found that low lying states are most affected by thermal effects, giant resonances being little affected in the range of temperatures here studied

  1. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    International Nuclear Information System (INIS)

    Simos, N.

    2011-01-01

    In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the

  2. Study on development of differential transformer for use in high-temperature environments

    International Nuclear Information System (INIS)

    Ara, Katsuyuki

    1983-11-01

    Today, in many fields of industrial science and technology, various efforts are being directed to the development of new technology aiming the technological inovation of the coming generation. Under these circumstances, new requirements are called for in instrumentation and measurement; one is the measurement at very severe environments such as high-temperature and high-pressure. Especially in the field of nuclear energy development, various kinds of measurements are needed under a high-temperature, high-pressure and high-radiation environments, and many sensors have been developed for such purposes. One of the most excellent heat-resisting sensors is the sensor based on and utilizing electromagnetic induction. Various electromagnetic sensors have been, therefore, developed and used in in-core environments of nuclear reactors. The author has been engaged in the development of differential transformers for use in in-core environments of Light Water Reactors: this paper compiles the results obtained through the development. (author)

  3. Effect of pairing in nuclear level density at low temperatures

    International Nuclear Information System (INIS)

    Rhine Kumar, A.K.; Modi, Swati; Arumugam, P.

    2013-01-01

    The nuclear level density (NLD) has been an interesting topic for researchers, due its importance in many aspects of nuclear physics, nuclear astrophysics, nuclear medicine, and other applied areas. The calculation of NLD helps us to understand the energy distribution of the excited levels of nuclei, entropy, specific heat, reaction cross sections etc. In this work the effect of temperature and pairing on level-density of the nucleus 116 Sn has been studied

  4. Database for nuclear-waste disposal for temperatures up to 3000C

    International Nuclear Information System (INIS)

    Phillips, S.L.; Silvester, L.F.

    1982-09-01

    A computerized database is compiled of evaluated thermodynamic data for aqueous species associated with nuclear waste storage. The data are organized as hydrolysis and formation constants; solubilities of oxides and hydroxides; and, as Nernstian potentials. More emphasis is on stability constants. Coeffficients are given to calculate stability constants at various ionic strengths and to high temperatures. Results are presented as tables for selected species including uranium, amorphous silica and actinides

  5. 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)

  6. 6th International Conference on Hard and Electromagnetic Probes of High-Energy Nuclear Collisions

    CERN Document Server

    2014-01-01

    One of the premier meetings in the field of high-energy nuclear physics, the Hard Probes conference series brings together the experimental and theoretical communities interested in the hard and electromagnetic observables related to nuclear matter at extreme temperatures and densities. Prior to the conference, the University of Cape Town will host a summer school for young physicists in the field. High energy nuclear physics focuses on the science of a trillion degrees. These temperatures were last seen in nature a microsecond after the Big Bang, but mankind recreates them thousands of times a second in particle accelerators such as CERN's Large Hadron Collider and BNL's Relativistic Heavy Ion Collider. At these temperatures, 100,000 times hotter than the center of the sun, the strong force is dominant, and we hope to learn about the fundamental and non-trivial emergent many-body dynamics of the quarks and gluons that make up 99% of the mass of the visible universe. We anticipate the usual format for the H...

  7. Temperature measurement in nuclear environment

    International Nuclear Information System (INIS)

    Degas, P.

    1986-12-01

    Some criterions, that the used sensors have to follow, are given together with the conditions they may encountered. They may be used in irradiation or safety test devices, in experiments concerning mock-up or plant element, or even in nuclear plants themselves. The most suitable sensor type are mentioned, with their characteristics and their fiability. Two use examples of temperature probes are given, chosen to illustrate two sensor types: thermocouples in Superphenix-1 and platinum resistance probes in research reactor Orphee [fr

  8. 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

  9. High Temperature Materials Interim Data Qualification Report

    International Nuclear Information System (INIS)

    Lybeck, Nancy

    2010-01-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim FY2010 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under NQA-1 guidelines, and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from two test series within the High Temperature Materials data stream have been entered into the NDMAS vault: (1) Tensile Tests for Sm (i.e., Allowable Stress) Confirmatory Testing - 1,403,994 records have been inserted into the NDMAS database. Capture testing is in process. (2) Creep-Fatigue Testing to Support Determination of Creep-Fatigue Interaction Diagram - 918,854 records have been processed and inserted into the NDMAS database. Capture testing is in process.

  10. Electrodeless, multi-megawatt reactor for room-temperature, lithium-6/deuterium nuclear reactions

    International Nuclear Information System (INIS)

    Drexler, J.

    1993-01-01

    This paper describes a reactor design to facilitate a room-temperature nuclear fusion/fission reaction to generate heat without generating unwanted neutrons, gamma rays, tritium, or other radioactive products. The room-temperature fusion/fission reaction involves the sequential triggering of billions of single-molecule, 6 LiD 'fusion energy pellets' distributed in lattices of a palladium ion accumulator that also acts as a catalyst to produce the molecules of 6 LiD from a solution comprising D 2 O, 6 LiOD with D 2 gas bubbling through it. The D 2 gas is the source of the negative deuterium ions in the 6 LiD molecules. The next step is to trigger a first nuclear fusion/fission reaction of some of the 6 LiD molecules, according to the well-known nuclear reaction: 6 Li + D → 2 4 He + 22.4 MeV. The highly energetic alpha particles ( 4 He nuclei) generated by this nuclear reaction within the palladium will cause shock and vibrations in the palladium lattices, leading to compression of other 6 LiD molecules and thereby triggering a second series of similar fusion/fission reactions, leading to a third series, and so on. The absorption of the kinetic energy in the palladium will, in turn, generate a continuous flow of heat into the heavy water carrier, which would be removed with a heat exchanger. (author)

  11. Conceptual design of a hydrogen production system by DME steam reforming and high-efficiency nuclear reactor technology

    International Nuclear Information System (INIS)

    Fukushima, Kimichika; Ogawa, Takashi

    2003-01-01

    Hydrogen is a potential alternative energy source and produced commercially by methane (natural gas) or LPG steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, since this process emits large amounts of CO 2 , replacement of the combustion heat source with a nuclear heat source for 773-1173 K processes has been proposed in order to eliminate these CO 2 emissions. This paper proposes a novel method of low-temperature nuclear hydrogen production by reforming dimethyl ether (DME) with steam produced by a low-temperature nuclear reactor at about 573 K. The authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573 K. By setting this low-temperature hydrogen production process at about 573K upstream from a turbine, it was found theoretically that the total energy utilization efficiency is about 50% and very high. By setting a turbine upstream of the hydrogen production plant, an overall efficiency of is 75% for an FBR and 76% for a supercritical-water cooled power reactor (SCPR). (author)

  12. High energy nuclear collisions: theory review

    International Nuclear Information System (INIS)

    Fries, Rainer J.

    2009-01-01

    Full text: High Energy Nuclear Collisions are studied at the Relativistic Heavy Ion Collider (RHIC) and, starting next year, also at the Large Hadron Collider (LHC) to study the formation and properties of quark gluon plasma (QGP). This effort is driven by the prediction that above a certain critical temperature quarks and gluons are deconfined. For the past ten years of running RHIC has performed marvelously. Data from RHIC has answered many initial questions, but it has also provided new, more challenging problems to understand the nature of quark gluon plasma and the dynamics of heavy ion collisions. In this talk I review some of the basic concepts of high energy nuclear collisions and quark gluon plasma formation. We also discuss some of the novel and open questions that we are faced with. We discuss recent predictions on properties of hot quantum chromodynamics, emerging signatures for the color glass condensate, the fascinating idea of local P and CP violation in QCD, as well as ongoing research on hard probes and electromagnetic signatures

  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. High-pressure, high-temperature magic angle spinning nuclear magnetic resonance devices and processes for making and using same

    Science.gov (United States)

    Hu, Jian Zhi; Hu, Mary Y.; Townsend, Mark R.; Lercher, Johannes A.; Peden, Charles H. F.

    2015-10-06

    Re-usable ceramic magic angle spinning (MAS) NMR rotors constructed of high-mechanic strength ceramics are detailed that include a sample compartment that maintains high pressures up to at least about 200 atmospheres (atm) and high temperatures up to about least about 300.degree. C. during operation. The rotor designs minimize pressure losses stemming from penetration over an extended period of time. The present invention makes possible a variety of in-situ high pressure, high temperature MAS NMR experiments not previously achieved in the prior art.

  16. Stable solid state reference electrodes for high temperature water chemistry

    International Nuclear Information System (INIS)

    Jayaweera, P.; Millett, P.J.

    1995-01-01

    A solid state electrode capable of providing a stable reference potential under a wide range of temperatures and chemical conditions has been demonstrated. The electrode consists of a zirconia or yttria-stabilized zirconia tube packed with an inorganic polymer electrolyte and a silver/silver chloride sensing element. The sensing element is maintained near room temperature by a passive cooling heat sink. The electrode stability was demonstrated by testing it in high temperature (280 C) aqueous solutions over extended periods of time. This reference electrode is useful in many applications, particularly for monitoring the chemistry in nuclear and fossil power plants

  17. Development of temperature profile sensor at high temporal and spatial resolution

    International Nuclear Information System (INIS)

    Takiguchi, Hiroki; Furuya, Masahiro; Arai, Takahiro

    2017-01-01

    In order to quantify thermo-physical flow field for the industrial applications such as nuclear and chemical reactors, high temporal and spatial measurements for temperature, pressure, phase velocity, viscosity and so on are required to validate computational fluid dynamics (CFD) and subchannel analyses. The paper proposes a novel temperature profile sensor, which can acquire temperature distribution in water at high temporal (a millisecond) and spatial (millimeter) resolutions. The devised sensor acquires electric conductance between transmitter and receiver wires, which is a function of temperature. The sensor comprise wire mesh structure for multipoint and simultaneous temperature measurement in water, which indicated that three-dimensional temperature distribution can be detected in flexible resolutions. For the demonstration of the principle, temperature profile in water was estimated according to pre-determined temperature calibration line against time-averaged impedance. The 16×16 grid sensor visualized fast and multi-dimensional mixing process of a hot water jet into a cold water pool. (author)

  18. Models of high energy nuclear collisions

    International Nuclear Information System (INIS)

    Glendenning, N.K.

    1978-06-01

    The discussion covers nuclear collisions at relativistic energies including classes of high energy nucleus--nucleus collisions, and the kinetics of a central collision; and the asymptotic hadron spectrum including known and unknown hadrons, the relevance of the spectrum and the means of its study, thermodynamics of hadronic matter, examples of hadronic spectra, the temperature, composition of the initial fireball and its expansion, isoergic expansion with no pre-freezeout radiation, isentropic expansion of the fireball, the quasi-dynamical expansion, and finally antinuclei, hypernuclei, and the quark phase. 28 references

  19. High-temperature spectroscopy for nuclear waste applications

    International Nuclear Information System (INIS)

    Grant, P.M.; Robouch, P.; Torres, R.A.; Silva, R.J.

    1991-10-01

    Instrumentation has been developed to perform uv-vis-nir absorbance measurements remotely and at elevated temperatures and pressures. Fiber-optic spectroscopy permits the interrogation of radioactive species within a glovebox enclosure at temperatures ranging from ambient to >100 degree C. Spectral shifts as a function of metal- ligand coordination are used to compute thermodynamic free energies of reaction by matrix regression analysis. Pr 3+ serves as a convenient analog for trivalent actinides without attendant radioactivity hazards, and recent results obtained from 20 degree--95 degree C with the Pr-acetate complexation system are presented. Preliminary experimentation on Am(3) hydrolysis is also described. 16 refs., 1 tab

  20. 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)

  1. 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)

  2. 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

  3. Hot nuclei: high temperatures, high angular momenta

    International Nuclear Information System (INIS)

    Guerreau, D.

    1991-01-01

    A review is made of the present status concerning the production of hot nuclei above 5 MeV temperature, concentrating mainly on the possible experimental evidences for the attainment of a critical temperature, on the existence of dynamical limitations to the energy deposition and on the experimental signatures for the formation of hot spinning nuclei. The data strongly suggest a nuclear disassembly in collisions involving very heavy ions at moderate incident velocities. Furthermore, hot nuclei seem to be quite stable against rotation on a short time scale. (author) 26 refs.; 12 figs

  4. 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

  5. Temperature distribution due to the heat generation in nuclear reactor shielding

    International Nuclear Information System (INIS)

    Torres, L.M.R.

    1985-01-01

    A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

  6. 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.''

  7. 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.'' .

  8. Study of nuclear level density parameter and its temperature dependence

    International Nuclear Information System (INIS)

    Nasrabadi, M. N.; Behkami, A. N.

    2000-01-01

    The nuclear level density ρ is the basic ingredient required for theoretical studies of nuclear reaction and structure. It describes the statistical nuclear properties and is expressed as a function of various constants of motion such as number of particles, excitation energy and angular momentum. In this work the energy and spin dependence of nuclear level density will be presented and discussed. In addition the level density parameter α will be extracted from this level density information, and its temperature and mass dependence will be obtained

  9. 6. Trilateral German-Russian-Ukrainian seminar on high-temperature superconductivity

    International Nuclear Information System (INIS)

    Aksenov, V.L.; Kornilov, E.I.

    1993-01-01

    The proceedings of the 6. Trilateral German-Russian-Ukrainian seminar on high-temperature superconductivity are reported. Nuclear methods (neutron diffraction, neutron polarization, Raman scattering, ESR) applied to study rare earth cuprates are described. The reports dealing with fundamental experimental studies and theoretical investigations are presented. The source materials for preparation of high-tc superconductors and the ways of their preparation, as well as their physico-chemical properties are considered

  10. Alpha-ray spectrometry at high temperature by using a compound semiconductor detector.

    Science.gov (United States)

    Ha, Jang Ho; Kim, Han Soo

    2013-11-01

    The use of conventional radiation detectors in harsh environments is limited by radiation damage to detector materials and by temperature constraints. We fabricated a wide-band gap semiconductor radiation detector based on silicon carbide. All the detector components were considered for an application in a high temperature environment like a nuclear reactor core. The radiation response, especially to alpha particles, was measured using an (241)Am source at variable operating voltages at room temperature in the air. The temperature on detector was controlled from 30°C to 250°C. The alpha-particle spectra were measured at zero bias operation. Even though the detector is operated at high temperature, the energy resolution as a function of temperature is almost constant within 3.5% deviation. Copyright © 2013 Elsevier Ltd. All rights reserved.

  11. Electrochemical cell and electrode designs for high-temperature/high-pressure kinetic measurements

    International Nuclear Information System (INIS)

    Nagy, Z.; Yonco, R.M.

    1987-05-01

    Many corrosion processes of interest to the nuclear power industry occur in high-temperature/high-pressure aqueous systems. The investigation of the kinetics of the appropriate electrode reactions is a serious experimental challenge, partially because of the high temperatures and pressures and partially because many of these reactions are very rapid, requiring fast relaxation measurements. An electrochemical measuring system is described which is suitable for measurements of the kinetics of fast electrode reactions at temperatures extending to at least 300 0 C and pressures to at least 10 MPa (100 atmospheres). The system includes solution preparation and handling equipment, the electrochemical cell, and several electrode designs. One of the new designs is a coaxial working electrode-counter electrode assembly; this electrode can be used with very fast-rising pulses, and it provides a well defined, repeatedly-polishable working surface. Low-impedance reference electrodes are also described, based on electrode concepts responding to the pH or the redox potential of the test solution. Additionally, a novel, long-life primary reference electrode design is reported, based on a modification of the external, pressure-balanced Ag/AgCl reference electrode

  12. Electrochemical cell and electrode designs for high-temperature/high-pressure kinetic measurements

    International Nuclear Information System (INIS)

    Nagy, Z.; Yonco, R.M.

    1988-01-01

    Many corrosion processes of interest to the nuclear power industry occur in high-temperature/high-pressure aqueous systems. The investigation of the kinetics of the appropriate electrode reactions is a serious experimental challenge, partially because of the high temperatures and pressures and partially because many of these reactions are very rapid, requiring fast relaxation measurements. An electrochemical measuring system is described which is suitable for measurements of the kinetics of fast electrode reactions at temperatures extending to at least 300 0 C and pressures to at least 10 MPa (100 atmospheres). The system includes solution preparation and handling equipment, the electrochemical cell, and several electrode designs. One of the new designs is a coaxial working electrode-counter electrode assembly; this electrode can be used with very fast-rising pulses, and it provides a well defined, repeatedly-polishable working surface. Low-impedance reference electrodes are also described, based on electrode concepts responding to the pH or the redox potential of the test solution. Additionally, a novel, long-life primary reference electrode design is reported, based on a modification of the external, pressure-balanced Ag/AgCl reference electrode

  13. High temperature structural design and R and Ds for heat transport system components of FBR 'Monju'

    International Nuclear Information System (INIS)

    Sumikawa, Masaharu; Nakagawa, Yukio; Fukuda, Yoshio; Sukegawa, Masayuki; Ishizaki, Tairo.

    1980-01-01

    The machines and equipments of cooling system for the fast breeder prototype reactor ''Monju'' are operated in creep temperature region, and the upper limit temperature to apply the domestic structural design standard for nuclear machines and equipment is exceeded, therefore the guideline for high temperature structural design is being drawn up, reflecting the results of recent research and development, by the Power Reactor and Nuclear Fuel Development Corp. and others. In order to obtain the basic data for the purpose, the tests on the high temperature characteristics of main structural members and structural elements were carried out, and eight kinds of the inelastic structural analysis program ''HI-EPIC'' series were developed, thus the fundamental technologies of structural desigh in non-linear region were established. Also in the non-linear region, enormous physical quantities must be evaluated, and in the design method based on real elastic analysis, many design diagrams must be employed, therefore for the purpose of improving the reliability of evaluation, the automatic evaluation program ''HI-TEP'' was developed, and preparation has been made for the design of actual machines. The high temperature structural design in ''Monju'', the development of inelastic structural analysis program and high temperature structural analysis evaluation program, and the development of high temperature structures and materials are described. (Kako, I.)

  14. A review of the high temperature creep in oxide nuclear fuels (I)

    International Nuclear Information System (INIS)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C. Y.

    1998-06-01

    Since the initial stage of fuel developmental until recently, considerable efforts have been extensively directed at studying the creep properties of uranium dioxide and its related phases largely due to the importance of their application to the reactor fuels. In this state-of-the-art report, the creep behavior and mechanisms of UO 2 and its related phases were reviewed and discussed in terms of experimental variables such as applied stress, temperature, microstructure and stoichiometry. The objective of this review is to obtain a complete understanding of the influences of these variables on the creep property and creep mechanism in these materials aiming at devising more proper methods for the improvement of the behavior. The database obtained from the results will be primarily utilized also, as the reference data for studying the creep behavior of UO 2 -based mixed oxide nuclear fuels. (author). 64 refs., 6 tabs., 25 figs

  15. An investigation of methods for neutron dose measurement in high temperature irradiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Kosako, Toshisou; Sugiura, Nobuyuki [Tokyo Univ. (Japan); Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan)] [and others

    2000-10-01

    The Japan Atomic Energy Research Institute (JAERI) has been conducting the innovative basic research on high temperature since 1994, which is a series of high temperature irradiation studies using the High Temperature Engineering Test Reactor (HTTR). 'The Task Group for Evaluation of Irradiation Dose under High Temperature Radiation' was founded in the HTTR Utilization Research Committee, which is the promoting body of the innovative basic research. The present report is a summary of investigation which has been made by the Task Group on the present status and subjects of research and development of neutron detectors in high temperature irradiation fields, in view of contributing to high temperature irradiation research using the HTTR. Detectors investigated here in the domestic survey are the following five kinds of in-core detectors: 1) small fission counter, 2) small fission chamber, 3) self-powered detector, 4) activation detector, and 5) optical fiber. In addition, the research and development status in Russia has been investigated. The present report will also be useful as nuclear instrumentation of high temperature gas-cooled reactors. (author)

  16. Lightweight, High-Temperature Radiator for Space Propulsion

    Science.gov (United States)

    Hyers, R. W.; Tomboulian, B. N.; Crave, Paul D.; Rogers, J. R.

    2012-01-01

    For high-power nuclear-electric spacecraft, the radiator can account for 40% or more of the power system mass and a large fraction of the total vehicle mass. Improvements in the heat rejection per unit mass rely on lower-density and higher-thermal conductivity materials. Current radiators achieve near-ideal surface radiation through high-emissivity coatings, so improvements in heat rejection per unit area can be accomplished only by raising the temperature at which heat is rejected. We have been investigating materials that have the potential to deliver significant reductions in mass density and significant improvements in thermal conductivity, while expanding the feasible range of temperature for heat rejection up to 1000 K and higher. The presentation will discuss the experimental results and models of the heat transfer in matrix-free carbon fiber fins. Thermal testing of other carbon-based fin materials including carbon nanotube cloth and a carbon nanotube composite will also be presented.

  17. Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

    International Nuclear Information System (INIS)

    Cabrera-Salcedo, A.

    2012-01-01

    During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

  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 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

  19. Performance Evaluation of Fabry-Perot Temperature Sensors in Nuclear Power Plant Measurements

    International Nuclear Information System (INIS)

    Liu Hanying; Miller, Don W.; Talnagi, Joseph W.

    2003-01-01

    The Fiso Fabry-Perot fiber-optic temperature sensor was selected for performance evaluation and for potential application in nuclear power plants because of its unique interferometric sensing mechanism and data-processing technique, and its commercial availability. It employs a Fizeau interferometer and a charge-coupled device array to locate the position of the maximum interference fringe intensity, which is directly related to the environmental temperature. Consequently, the basic sensing mechanism is independent of the absolute transmitted light intensity, which is the most likely parameter to be affected by external harsh environments such as nuclear irradiation, high pressure/temperature, and cyclical vibration.This paper reports research on the performance of two Fiso Fabry-Perot temperature sensors in environmental conditions expected in nuclear power plants during both normal and abnormal (i.e., accident) conditions. The environmental conditions simulated in this paper include gamma-only ( 60 Co) irradiation, pressure/temperature environmental transient, and mixed neutron/gamma field, respectively.The first sensor exhibited no failure or degradation in performance during and following gamma-only irradiation in which a total dose of 15 kGy was delivered at a dose rate of 2.5 kGy/h. Following gamma irradiation, this sensor was then tested for 10.75 days in a thermohydraulic environment prescribed by the Institute of Electrical and Electronics Engineers IEEE323-1983. Intermittent behavior was observed throughout the latter portions of this test, and degradation in performance occurred after the test. Visual evaluation after opening the sensor head indicated that the internal welding methodology was the primary contributor to the observed behavior during this test. Further consultation with the vendor shows that the robustness and reliability of Fiso sensors can be substantially improved by modifying the internal welding methods.The second Fiso temperature

  20. The elevated temperature and thermal shock fracture toughnesses of nuclear pressure vessel steel

    International Nuclear Information System (INIS)

    Hirano, Kazumi; Kobayashi, Hideo; Nakazawa, Hajime; Nara, Atsushi.

    1979-01-01

    Thermal shock experiments were conducted on nuclear pressure vessel steel A533 Grade B Class 1. Elastic-plastic fracture toughness tests were carried out within the same high temperature range of the thermal shock experiment and the relation between stretched zone width, SZW and J-integral was clarified. An elastic-plastic thermal shock fracture toughness value. J sub(tsc) was evaluated from a critical value of stretched zone width, SZW sub(tsc) at the initiation of thermal shock fracture by using the relation between SZW and J. The J sub(tsc) value was compared with elastic-plastic fracture toughness values, J sub( ic), and the difference between the J sub(tsc) and J sub( ic) values was discussed. The results obtained are summarized as follows; (1) The relation between SZW and J before the initiation of stable crack growth in fracture toughness test at a high temperature can be expressed by the following equation regardless of test temperature, SZW = 95(J/E), where E is Young's modulus. (2) Elevated temperature fracture toughness values ranging from room temperature to 400 0 C are nearly constant regardless of test temperature. It is confirmed that upper shelf fracture toughness exists. (3) Thermal shock fracture toughness is smaller than elevated temperature fracture toughness within the same high temperature range of thermal shock experiment. (author)

  1. 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.

  2. 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

  3. High temperature, radiation hardened electronics for application to nuclear power plants

    International Nuclear Information System (INIS)

    Gover, J.E.

    1980-01-01

    Electronic circuits were developed and built at Sandia for many aerospace and energy systems applications. Among recent developments were high temperature electronics for geothermal well logging and radiation hardened electronics for a variety of aerospace applications. Sandia has also been active in technology transfer to commercial industry in both of these areas

  4. 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

  5. High-Temperature Piezoelectric Sensing

    Directory of Open Access Journals (Sweden)

    Xiaoning Jiang

    2013-12-01

    Full Text Available Piezoelectric sensing is of increasing interest for high-temperature applications in aerospace, automotive, power plants and material processing due to its low cost, compact sensor size and simple signal conditioning, in comparison with other high-temperature sensing techniques. This paper presented an overview of high-temperature piezoelectric sensing techniques. Firstly, different types of high-temperature piezoelectric single crystals, electrode materials, and their pros and cons are discussed. Secondly, recent work on high-temperature piezoelectric sensors including accelerometer, surface acoustic wave sensor, ultrasound transducer, acoustic emission sensor, gas sensor, and pressure sensor for temperatures up to 1,250 °C were reviewed. Finally, discussions of existing challenges and future work for high-temperature piezoelectric sensing are presented.

  6. Peculiar Features of Thermal Aging and Degradation of Rapidly Quenched Stainless Steels under High-Temperature Exposures

    Science.gov (United States)

    Shulga, A. V.

    2017-12-01

    This article presents the results of comparative studies of mechanical properties and microstructure of nuclear fuel tubes and semifinished stainless steel items fabricated by consolidation of rapidly quenched powders and by conventional technology after high-temperature exposures at 600 and 700°C. Tensile tests of nuclear fuel tube ring specimens of stainless austenitic steel of grade AISI 316 and ferritic-martensitic steel are performed at room temperature. The microstructure and distribution of carbon and boron are analyzed by metallography and autoradiography in nuclear fuel tubes and semifinished items. Rapidly quenched powders of the considered steels are obtained by the plasma rotating electrode process. Positive influence of consolidation of rapidly quenched powders on mechanical properties after high-temperature aging is confirmed. The correlation between homogeneous distribution of carbon and boron and mechanical properties of the considered steel is determined. The effects of thermal aging and degradation of the considered steels are determined at 600°C and 700°C, respectively.

  7. High-temperature, high-pressure bonding of nested tubular metallic components

    International Nuclear Information System (INIS)

    Quinby, T.C.

    1980-01-01

    This invention is a tool for effecting high-temperature, high compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hotpress evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity

  8. High-temperature, high-pressure bonding of nested tubular metallic components

    Science.gov (United States)

    Quinby, T.C.

    A tool is described for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators. The target assembly comprising a uranum foil and an aluninum-alloy substrate. The tool is composed of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

  9. Proceeding of the 7. Seminar on Technology and Safety of Nuclear Power Plants and Nuclear Facilities

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Antariksawan, Anhar R.; Soetrisnanto, Arnold Y; Jujuratisbela, Uju; Aziz, Ferhat; Su'ud, Zaki; Suprawhardana, M. Salman

    2002-02-01

    The seventh proceedings of seminar safety and technology of nuclear power plant and nuclear facilities, held by National Nuclear Energy Agency. The Aims of seminar is to exchange and disseminate information about safety and nuclear Power Plant Technology and Nuclear Facilities consist of technology; high temperature reactor and application for national development sustain able and high technology. This seminar level all aspects technology, Power Reactor research reactor, high temperature reactor and nuclear facilities. The article is separated by index

  10. Controllers for high-performance nuclear fusion plasmas

    NARCIS (Netherlands)

    Baar, de M.R.

    2012-01-01

    A succesful nuclear fusion reactor will confine plasma at hig temperatures and densities, with low thermal losses. The workhorse of the nuclear fusion community is the tokamak, a toroidal device in which plasmas are confined by poloidal and toroidal magnetic fields. Ideally, the confirming magnetic

  11. 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)

  12. 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.

  13. The problems of using a high-temperature sodium coolant in nuclear power plants for the production of hydrogen and other innovative applications

    Science.gov (United States)

    Sorokin, A. P.; Alexeev, V. V.; Kuzina, Ju. A.; Konovalov, M. A.

    2017-11-01

    The intensity of the hydrogen sources arriving from the third contour of installation in second in comparison with the hydrogen sources on NPP BN-600 increases by two - three order at using of high-temperature nuclear power plants with the sodium coolant (HT-NPP) for drawing of hydrogen and other innovative applications (gasification and a liquefaction of coal, profound oil refining, transformation of biomass to liquid fuel, in the chemical industry, metallurgy, the food-processing industry etc.). For these conditions basic new technological solutions are offered. The main condition of their implementation is raise of hydrogen concentration in the sodium coolant on two - three order in comparison with the modern NPP, in a combination to hydrogen removal from sodium and its pumping out through membranes from vanadium or niobium. The researches with use diffusive model have shown possibility to expel a casium inflow in sodium through a leakproof shell of fuel rods if vary such parameters as a material of fuel rods shell, its thickness and maintenance time at design of fuel rods for high-temperature NPP. However maintenance of high-temperature NPP in the presence of casium in sodium is inevitable at loss of leakproof of a fuel rods shell. In these conditions for minimisation of casium diffusion in structural materials it is necessary to provide deep clearing of sodium from cesium.

  14. Industrial Qualification Process for Optical Fibers Distributed Strain and Temperature Sensing in Nuclear Waste Repositories

    Directory of Open Access Journals (Sweden)

    S. Delepine-Lesoille

    2012-01-01

    Full Text Available Temperature and strain monitoring will be implemented in the envisioned French geological repository for high- and intermediate-level long-lived nuclear wastes. Raman and Brillouin scatterings in optical fibers are efficient industrial methods to provide distributed temperature and strain measurements. Gamma radiation and hydrogen release from nuclear wastes can however affect the measurements. An industrial qualification process is successfully proposed and implemented. Induced measurement uncertainties and their physical origins are quantified. The optical fiber composition influence is assessed. Based on radiation-hard fibers and carbon-primary coatings, we showed that the proposed system can provide accurate temperature and strain measurements up to 0.5 MGy and 100% hydrogen concentration in the atmosphere, over 200 m distance range. The selected system was successfully implemented in the Andra underground laboratory, in one-to-one scale mockup of future cells, into concrete liners. We demonstrated the efficiency of simultaneous Raman and Brillouin scattering measurements to provide both strain and temperature distributed measurements. We showed that 1.3 μm working wavelength is in favor of hazardous environment monitoring.

  15. Neutron studies of nuclear magnetism at ultralow temperature

    DEFF Research Database (Denmark)

    Siemensmeyer, K.; Clausen, K.N.; Lefmann, K.

    1997-01-01

    Nuclear magnetic order in copper and silver has been investigated by neutron diffraction. Antiferromagnetic order is observed in these simple, diamagnetic metals at temperatures below 50 nK and 560 pK, respectively. Both crystallize in the FCC-symmetry which is fully frustrated for nearest...

  16. 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

  17. A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Cho, M. S.; Choo, K. N.; Shin, Y. T.; Sohn, J. M.; Park, S. J.; Kim, B. G

    2008-01-15

    1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

  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. 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.)

  20. High temperature compression tests performed on doped fuels

    International Nuclear Information System (INIS)

    Duguay, C.; Mocellin, A.; Dehaudt, P.; Fantozzi, G.

    1997-01-01

    The use of additives of corundum structure M 2 O 3 (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO 2 doped with these oxides has been investigated and compared with that of pure UO 2 with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.)

  1. Highly efficient high temperature electrolysis

    DEFF Research Database (Denmark)

    Hauch, Anne; Ebbesen, Sune; Jensen, Søren Højgaard

    2008-01-01

    High temperature electrolysis of water and steam may provide an efficient, cost effective and environmentally friendly production of H-2 Using electricity produced from sustainable, non-fossil energy sources. To achieve cost competitive electrolysis cells that are both high performing i.e. minimum...... internal resistance of the cell, and long-term stable, it is critical to develop electrode materials that are optimal for steam electrolysis. In this article electrolysis cells for electrolysis of water or steam at temperatures above 200 degrees C for production of H-2 are reviewed. High temperature...... electrolysis is favourable from a thermodynamic point of view, because a part of the required energy can be supplied as thermal heat, and the activation barrier is lowered increasing the H-2 production rate. Only two types of cells operating at high temperature (above 200 degrees C) have been described...

  2. High-temperature process heat applications with an HTGR

    International Nuclear Information System (INIS)

    Quade, R.N.; Vrable, D.L.

    1980-04-01

    An 842-MW(t) HTGR-process heat (HTGR-PH) design and several synfuels and energy transport processes to which it could be coupled are described. As in other HTGR designs, the HTGR-PH has its entire primary coolant system contained in a prestressed concrete reactor vessel (PCRV) which provides the necessary biological shielding and pressure containment. The high-temperature nuclear thermal energy is transported to the externally located process plant by a secondary helium transport loop. With a capability to produce hot helium in the secondary loop at 800 0 C (1472 0 F) with current designs and 900 0 C (1652 0 F) with advanced designs, a large number of process heat applications are potentially available. Studies have been performed for coal liquefaction and gasification using nuclear heat

  3. Comprehensive study of temperature anomalies on of the former Semipalatinsk nuclear test site territory

    International Nuclear Information System (INIS)

    Subbotin, S.B.; Lukashenko, S.N.; Dmitropavlenko, V.N.; Ajdarkhanov, A.O.; Duchkov, A.D.; Kazantsev, S.A.

    2005-01-01

    In 1997 by the space images data in the Semipalatinsk test site area a mysterious anomaly thermal zone with square about 20 thousand sq. km. with soil temperature 10-15 degrees above than on the adjacent areas was found. The results of 1996-1999 observation confirm the presence of steady temperature anomalies. A number of scientists are suggesting that the increased temperature zones are related with conducted nuclear tests. These temperature anomalies related with objects of nuclear explosions conduction and its have limited distribution the spatially attached to nuclear explosions cavities. Anomalies are the sequent of residual manifestation of long-time geothermal activity in the underground nuclear explosions epicenters. In 2001 in the frameworks of joint program 'Comprehensive study of thermal anomalies on the territory of the former Semipalatinsk test site' the direct measurements of soils on the five sections which were selected by the results of space images

  4. Immobilisation of high level nuclear reactor wastes in SYNROC

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W; Major, A [Australian National Univ., Canberra. Inst. of Advanced Studies

    1979-03-15

    It is stated that the elements occurring in high-level nuclear reactor wastes can be safely immobilised by incorporating them within the crystal lattices of the constituent minerals of a synthetic rock (SYNROC). The preferred form of SYNROC can accept up to 20% of high level waste calcine to form dilute solid solutions. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20 to 2,000 Myr whilst immobilising the same elements present in nuclear wastes. SYNROC is unaffected by leaching for 24 hours in pure water or 10 wt % NaCl solution at high temperatures and pressure whereas borosilicate glasses completely decompose in a few hours in much less severe hydrothermal conditions. The combination of these leaching results with the geological evidence of long-term stability indicates that SYNROC would be vastly superior to glass in its capacity to safely immobilise nuclear wastes, when buried in a suitable geological repository. A dense, compact, mechanically strong form of SYNROC suitable for geological disposal can be produced by a process as economical as that which incorporates radioactive waste in borosilicate glasses.

  5. High temperature creep strength of Advanced Radiation Resistant Oxide Dispersion Strengthened Steels

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, high temperature strength of advanced radiation resistance ODS steel was investigated for the core structural material of next generation nuclear systems. ODS martensitic steel was designed to have high homogeneity, productivity and reproducibility. Mechanical alloying, hot isostactic pressing and hot rolling processes were employed to fabricate the ODS steels, and creep rupture test as well as tensile test were examined to investigate the behavior at high temperatures. ODS steels were fabricated by a mechanical alloying and hot consolidation processes. Mechanical properties at high temperatures were investigated. The creep resistance of advanced radiation resistant ODS steels was more superior than those of ferritic/ martensitic steel, austenitic stainless steel and even a conventional ODS steel.

  6. 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.

  7. 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.

  8. Evaluation of Dynamic Reversible Chemical Energy Storage with High Temperature Electrolysis

    OpenAIRE

    McVay, Derek Joseph

    2017-01-01

    Renewable power generation is intermittent and non-dispatchable, but is steadily increasing in penetration due to lower costs associated with installation and demand for clean power generation. Without significant energy storage available to a grid with high renewable penetration, a mismatch between the load and the power available can. Furthermore, advanced high temperature nuclear reactors offer clean power generation, but only at a baseload operation scenario due to the significant thermal...

  9. High temperature materials and mechanisms

    CERN Document Server

    2014-01-01

    The use of high-temperature materials in current and future applications, including silicone materials for handling hot foods and metal alloys for developing high-speed aircraft and spacecraft systems, has generated a growing interest in high-temperature technologies. High Temperature Materials and Mechanisms explores a broad range of issues related to high-temperature materials and mechanisms that operate in harsh conditions. While some applications involve the use of materials at high temperatures, others require materials processed at high temperatures for use at room temperature. High-temperature materials must also be resistant to related causes of damage, such as oxidation and corrosion, which are accelerated with increased temperatures. This book examines high-temperature materials and mechanisms from many angles. It covers the topics of processes, materials characterization methods, and the nondestructive evaluation and health monitoring of high-temperature materials and structures. It describes the ...

  10. Study of thermocouples for control of high temperatures

    International Nuclear Information System (INIS)

    Villamayor, M.

    1966-12-01

    Previous works have shown that the tungsten-rhenium alloys thermocouples were a good instrument for control of high temperatures. From its, the author has studied the W/W 26 per cent and W 5 per cent Re/W 26 per cent Re french manufactured thermocouples and intended for control of temperatures in nuclear reactors until 2300 deg. C. In 'out-pile' study he determines the general characteristics of these thermocouples: average calibration curves, thermal shocks influence, response times, and alloys allowing the cold source compensation. The evolution of these thermocouples under thermal neutron flux has been determined by 'in-pile' study. The observations have led the author to propose a new type of thermocouples settled of molybdenum-columbium alloys. (author) [fr

  11. Frictional characteristics of silicon graphite lubricated with water at high pressure and high temperature

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Eun Hyun; Park, Jin Seok; Kim, Jong In

    2001-01-01

    Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss and wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics

  12. Molybdenum Disilicide Oxidation Kinetics in High Temperature Steam

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Stephen Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-07

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials, or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi2 was identified based on its high temperature oxidation resistance in O2 atmospheres (e.g. air and “wet air”). However, its behavior in H2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670

  13. High temperature refrigerator

    International Nuclear Information System (INIS)

    Steyert, W.A. Jr.

    1978-01-01

    A high temperature magnetic refrigerator is described which uses a Stirling-like cycle in which rotating magnetic working material is heated in zero field and adiabatically magnetized, cooled in high field, then adiabatically demagnetized. During this cycle the working material is in heat exchange with a pumped fluid which absorbs heat from a low temperature heat source and deposits heat in a high temperature reservoir. The magnetic refrigeration cycle operates at an efficiency 70% of Carnot

  14. Optimum development temperature and duration for nuclear plate

    International Nuclear Information System (INIS)

    Nagoshi, Chieko.

    1975-01-01

    Sakura 100 μm thick nuclear plates have been employed to determine optimum temperature and duration of the Amidol developer for low energy protons (Ep 0 C were tried for periods of 15--35 min. For Ep 0 C and for development time less than 30 min. (auth.)

  15. 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

  16. 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

  17. Performance test of ex-core high temperature and high pressure water loop test equipment (Contract research)

    International Nuclear Information System (INIS)

    Nakano, Hiroko; Uehara, Toshiaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nakamura, Jinichi; Matsui, Yoshinori; Tsuchiya, Kunihiko

    2016-03-01

    In Japan Atomic Energy Agency, we started research and development so as to monitor the situations in the Nuclear Plant Facilities during a severe accident, such as a radiation-resistant monitoring camera, a radiation-resistant transmission system for conveying the in-core information, and a heat-resistant signal cable. As a part of developments of the heat-resistant signal cable, we prepared ex-core high-temperature and high-pressure water loop test equipment, which can simulate the conditions of BWRs and PWRs, for evaluating reliability and properties of sheath materials of the cable. This equipment consists of autoclave, water conditioning tank, high-pressure metering pump, preheater, heat exchanger and water purification equipment, etc. This report describes the basic design and the performance test results of ex-core high-temperature and high-pressure water loop test equipment. (author)

  18. Temperature uniformity mapping in a high pressure high temperature reactor using a temperature sensitive indicator

    NARCIS (Netherlands)

    Grauwet, T.; Plancken, van der I.; Vervoort, L.; Matser, A.M.; Hendrickx, M.; Loey, van A.

    2011-01-01

    Recently, the first prototype ovomucoid-based pressure–temperature–time indicator (pTTI) for high pressure high temperature (HPHT) processing was described. However, for temperature uniformity mapping of high pressure (HP) vessels under HPHT sterilization conditions, this prototype needs to be

  19. Proceedings of the 9. National Seminar on Technology and Safety of Nuclear Power Plants and Nuclear Facilities

    International Nuclear Information System (INIS)

    Antariksawan, Anhar R.; Soetrisnanto, Arnold Y; Aziz, Ferhat; Untoro, Pudji; Su'ud, Zaki; Zarkasi, Amin Santoso; Lasman, As Natio

    2003-08-01

    The ninth proceedings of seminar safety and technology of nuclear power plant and nuclear facilities held by National Nuclear Energy Agency and PLN-JTK. The aims of seminar is to exchange and disseminate information about Safety and Nuclear Power Plant Technology and Nuclear Facilities consist of Technology High Temperature Reactor and Application for National Development Sustainable and High Technology. This seminar cover all aspects Technology, Power Reactor, Research Reactor High Temperature Reactor and Nuclear Facilities. There are 20 articles have separated index

  20. Certification of temperature measuring techniques at thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Preobrazhenskij, V.P.; Strigina, L.A.

    1980-01-01

    Necessity for metrological certification of temperature measurement techniques (TMT) at thermal and nuclear energy plants is grounded. An order of TMT certification is stated and formulae for determining the accuracy of temperature measurements by the thermoelectric method are given. It is concluded that through there are also statistical characteristics of errors of a number of measurement properties, it is necessary to carry on statistical investigations into errors of thermoelectrode extending wires, planimeters, measurement conditions. Such kind investigation technigues have been developed. Besides, it is necessary to regulate a uniform approach to the usage of statistical characteristics of errors of means and conditions of measurements to minimize volume of work for the personnel of thermal and nuclear energy plants and provide reliable estimates of temperature measurement errors

  1. Low cycle fatigue and creep fatigue behavior of alloy 617 at high temperature

    International Nuclear Information System (INIS)

    Cabet, Celine; Carroll, Laura; Wright, Richard

    2013-01-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger (IHX) application of the very high temperature nuclear reactor (VHTR), expected to have an outlet temperature as high as 950 C. Acceptance of Alloy 617 in Section III of the ASME Code for nuclear construction requires a detailed understanding of the creep-fatigue behavior. Initial creep-fatigue work on Alloy 617 suggests a more dominant role of environment with increasing temperature and/or hold times evidenced through changes in creep-fatigue crack growth mechanisms and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle fatigue specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens and the addition of a hold time at peak tensile strain degraded the cycle life. This suggests that creep-fatigue interaction occurs and that the environment may be partially responsible for accelerating failure. (authors)

  2. High temperature water chemistry monitoring

    International Nuclear Information System (INIS)

    Aaltonen, P.

    1992-01-01

    Almost all corrosion phenomena in nuclear power plants can be prevented or at least damped by water chemistry control or by the change of water chemistry control or by the change of water chemistry. Successful water chemistry control needs regular and continuous monitoring of such water chemistry parameters like dissolved oxygen content, pH, conductivity and impurity contents. Conventionally the monitoring is carried out at low pressures and temperatures, which method, however, has some shortcomings. Recently electrodes have been developed which enables the direct monitoring at operating pressures and temperatures. (author). 2 refs, 5 figs

  3. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  4. 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.)

  5. High temperature slagging incinerator for TRU-waste treatment

    International Nuclear Information System (INIS)

    Van De Voorde, N.; Hennart, D.; Gijbels, J.; Mergan, L.

    1984-01-01

    Since 1974 the Belgian Nuclear Study Center (SCK/CEN) at Mol, with the support of the European Communities, has developed an ''integral'' system for the treatment and the conditioning of radioactive contaminated wastes. The system converts directly, at high temperature (1500 0 C), mixtures of combustibles (paper, plastics, rubber etc.) and non-combustibles (metals, soil, sludge, concrete.) contaminated with transuranium elements as well as beta-gamma emitting isotopes, into a chemically inert and physically stable slag. More than 4000 hours of successful operation, with wide variety of simulated waste composition as well as real waste, have confirmed the safe operability of the high temperature sl'Gging incinerator and the connected installations, such as sorting cells, waste shredder, off-gas purification train, slag extraction system, remoted control, and the alpha-containment building. During the fall of 1983, a final confirmation of the performance of the installation was given by the successful accomplishment of an incineration campaign of 16 to 17 tons of simulated solid plutonium contaminated wastes

  6. IAEA Activities in Nuclear High Temperature Heat for Industrial Processes

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2017-01-01

    IAEA activities to support Member States: Information Exchange; Modelling and Simulations; Development of Methodologies; Safety; Technology Support; Education and Training; Knowledge Preservation. Assist MSs with national nuclear programmes; Support innovations in nuclear power deployment; Facilitate and assist international R&D collaborations. Interest in HTGR technology • The IAEA activities in the area of HTGR are guided by the recommendations of the TWG-GCRs – Currently 14 members: China, France, Germany, Indonesia, Japan, Korea (Rep. of), Netherlands, Russian Federation, South Africa, Switzerland, Turkey, Ukraine, United Kingdom, United States of America – 3 International Organizations: OECD/NEA, European Commission, Gen-IV Forum. – 2 new members in 2017: Poland and Singapore. Meetings • Meet every 24 months • Next meeting: 30 October – 1 November 2017 • Other Member states with some activities on HTGRs – Kazakhstan – history of close cooperation with Japan – Saudi Arabia – feasibility study for HTGRs to provide heat for the petro-chemical industry – Canada – three HTR designs under consideration in the nuclear regulator pre-licensing vendor design reviews

  7. Iron-free moving coil high temperature displacement transducer

    Energy Technology Data Exchange (ETDEWEB)

    Grindrod, A

    1976-07-01

    A unique, iron free, moving coil linear displacement transducer system is described, which is suitable for continuously monitoring linear movements, at varying temperatures up to 750/sup 0/C, in operational nuclear reactors. Although this device has been primarily developed for Advanced Gas Cooled Reactor Systems, it also has uses where long term measurements on conventional high temperature plant are required. Furthermore it could be particularly useful in material creep laboratories where precise linear changes in specimen length need to be monitored at elevated temperatures, over several years. Since individual transducer installations demand specific mounting arrangements to suit particular component geometries, evaluations have been made only on standard operational modules or capsules which are designed for containment in a range of housing or fixtures to suit particular applications. The behaviour of these devices has been studied at temperatures up to 750/sup 0/C for periods of over 10,000 h. An evaluation is also included of a commercially designed sensor assembly employing the same principle, for monitoring the boiler-shield wall movement at Hinkley Point 'B' AGR Station.

  8. High-temperature laser induced spectroscopy in nuclear steam generators

    International Nuclear Information System (INIS)

    Allmon, W.E.; Berthold, J.W.

    1990-01-01

    This patent describes an apparatus for conducting optical spectroscopy in a hostile environment. It comprises: a source of high intensity light; an optical fiber connected to the source of high intensity light for transmitting light therefrom. The optical fiber having an end for discharging light onto a material to be spectroscopically analyzed; a sheath defining a space around at least a part of the optical fiber carrying the end of the optical fiber for shielding the optical fiber from the hostile environment; a window in the sheath for closing the space and for passing light transmitted through the end of the optical fiber out of the sheath; light detector means for detecting and spectroscopically analyzing emitted light from the material; an optical fiber means for transmitting the emitted light from the material to the light detector means; a standardization module for containing a sample having a known composition and being exposed to known temperature and pressure conditions; an additional optical fiber connected to the module for transmitting light to the sample in the module; multiplexer means; and additional optical fiber means for returning light from the module to the detector through the multiplexer means

  9. Annular air space effects on nuclear waste canister temperatures in a deep geologic waste repository

    International Nuclear Information System (INIS)

    Lowry, W.E.; Cheung, H.; Davis, B.W.

    1980-01-01

    Air spaces in a deep geologic repository for nuclear high level waste will have an important effect on the long-term performance of the waste package. The important temperature effects of an annular air gap surrounding a high level waste canister are determined through 3-D numerical modeling. Air gap properties and parameters specifically analyzed and presented are the air gap size, surfaces emissivity, presence of a sleeve, and initial thermal power generation rate; particular emphasis was placed on determining the effect of these variables have on the canister surface temperature. Finally a discussion based on modeling results is presented which specifically relates the results to NRC regulatory considerations

  10. Results and future plans for the innovative basic research on high temperature engineering

    International Nuclear Information System (INIS)

    2001-05-01

    The High Temperature Engineering Test Reactor (HTTR) is under the rise-to-power stage at the Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). This reactor is aimed not only at establishment of the infrastructural technology on high temperature gas-cooled reactor and its upgrading, but also at promotion of the innovative basic research on high temperature engineering. The research is a series of innovative high-temperature irradiation studies, making the best use of the characteristic of the HTTR that it provides a very wide irradiation space at high temperatures. The JAERI has been conducting preliminary tests of the innovative research since 1994, in collaboration with universities and other research institutes, in the fields of 1) new materials development, 2) high temperature radiation chemistry and fusion-related research, and 3) high temperature irradiation techniques and other nuclear research. The HTTR Utilization Research Committee has been examining the results and methodology of the preliminary tests and the future plans, as well as examining the preparatory arrangements of facilities for the HTTR irradiation and post-irradiation examinations. This report presents a summary of results of the preliminary tests and preparatory arrangements for about seven years, together with an outline of the future plans. (author)

  11. Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

    Science.gov (United States)

    Waseem, Owais Ahmed; Jeong, Jong-Ryul; Park, Byong-Guk; Maeng, Cheol-Soo; Lee, Myoung-Goo; Ryu, Ho Jin

    2017-11-01

    The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

  12. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels

  13. 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)

  14. 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

  15. 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.)

  16. High temperature compression tests performed on doped fuels

    Energy Technology Data Exchange (ETDEWEB)

    Duguay, C.; Mocellin, A.; Dehaudt, P. [Commissariat a l`Energie Atomique, CEA Grenoble (France); Fantozzi, G. [INSA Lyon - GEMPPM, Villeurbanne (France)

    1997-12-31

    The use of additives of corundum structure M{sub 2}O{sub 3} (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO{sub 2} doped with these oxides has been investigated and compared with that of pure UO{sub 2} with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.) 5 refs.

  17. Charge redistribution and properties of high-temperature superconductors

    International Nuclear Information System (INIS)

    Khomskii, D.I.; Kusmartsev, F.V.

    1992-01-01

    We show that in high-T c superconductors (HTSC) with two groups of electrons (e.g., holes in CuO 2 planes and in a ''reservoir'') there should exist a charge redistribution with the temperature: the hole concentration N h in ''active'' superconducting CuO 2 planes increases below T c . This effect may explain structural changes such as the shift of the apical oxygen atom, anomalous thermal expansion, the shift of nuclear quadrupole resonance lines, the change of the positron lifetime, and the modification of the ion channeling below T c . Some other possible consequences of the charge redistribution (the modification of the temperature dependence of a gap Δ and of the ratio 2Δ 0 /T c , the phenomena at a contact of HTSC with normal metals and semiconductors) are discussed

  18. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

    1994-08-01

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  19. LVDT Development for High Temperature Irradiation Test and Application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Ban, Chae Min; Choo, Kee Nam; Jun, Byung Hyuk [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The LVDT (Linear Variable Differential Transformer) is used to measure the elongation and pressure of a nuclear fuel rod, or the creep and fatigue of the material during a reactor irradiation test. This device must be a radiation-resistant LVDT for use in a research reactor. Norway Halden has LVDTs for an irradiation test by the own development and commercialized. But Halden's LVDTs have limited the temperature of the use until to 350 .deg. C. So, KAERI has been developing a new LVDT for high temperature irradiation test. This paper describes the design of a LVDT, the fabrication process of a LVDT, and the result of the performance test. The designed LVDT uses thermocouple cable for coil wire material and one MI cable as signal cable. This LVDT for a high temperature irradiation test can be used until a maximum of 900 .deg. C. Welding is a very important factor for the fabrication of an LVDT. We are using a 150W fiber laser welding system that consists of a welding head, monitoring vision system and rotary index.

  20. Effect of high temperature or fire on heavy weight concrete properties

    International Nuclear Information System (INIS)

    Sakr, K.; EL-Hakim, E.

    2005-01-01

    Temperature plays an important role in the use of concrete for shielding nuclear reactors. In the present work, the effect of different durations (1, 2 and 3 h) of high temperatures (250, 500, 750 and 950 deg. C) on the physical, mechanical and radiation properties of heavy concrete was studied. The effect of fire fitting systems on concrete properties was investigated. Results showed that ilmenite concrete had the highest density, modulus of elasticity and lowest absorption percent, and it had also higher values of compressive, tensile, bending and bonding strengths than gravel or baryte concrete. Ilmenite concrete showed the highest attenuation of transmitted gamma rays. Firing (heating) exposure time was inversely proportional to mechanical properties of all types of concrete. Ilmenite concrete was more resistant to elevated temperature. Foam or air proved to be better than water as a cooling system in concrete structure exposed to high temperature because water leads to a big damage in concrete properties

  1. Carbon-13 Labeling Used to Probe Cure and Degradation Reactions of High- Temperature Polymers

    Science.gov (United States)

    Meador, Mary Ann B.; Johnston, J. Christopher

    1998-01-01

    High-temperature, crosslinked polyimides are typically insoluble, intractible materials. Consequently, in these systems it has been difficult to follow high-temperature curing or long-term degradation reactions on a molecular level. Selective labeling of the polymers with carbon-13, coupled with solid nuclear magnetic resonance spectrometry (NMR), enables these reactions to be followed. We successfully employed this technique to provide insight into both curing and degradation reactions of PMR-15, a polymer matrix resin used extensively in aircraft engine applications.

  2. 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

  3. Assessment of very high-temperature reactors in process applications. Appendix II. VHTR process heat application studies

    International Nuclear Information System (INIS)

    Jones, J.E.; Gambill, W.R.; Cooper, R.H.; Fox, E.C.; Fuller, L.C.; Littlefield, C.C.; Silverman, M.D.

    1977-06-01

    A critical review is presented of the technology and economics for coupling a very high-temperature gas-cooled reactor to a variety of process applications. It is concluded that nuclear steam reforming of light hydrocarbons for coal conversion could be a near-term alternative and that direct nuclear coal gasification could be a future consideration. Thermochemical water splitting appears to be more costly and its availability farther in the future than the coal-conversion systems. Nuclear steelmaking is competitive with the direct reduction of iron ore from conventional coal-conversion processes but not competitive with the reforming of natural gas at present gas prices. Nuclear process heat for petroleum refining, even with the necessary backup systems, is competitive with fossil energy sources. The processing with nuclear heat of oil shale and tar sands is of marginal economic importance. An analysis of peaking power applications using nuclear heat was also made. It is concluded that steam reforming methane for energy storage and production of peaking power is not a viable economic alternative, but that energy storage with a high-temperature heat transfer salt (HTS) is competitive with conventional peaking systems. An examination of the materials required in process heat exchangers is made

  4. Perry Nuclear Power Plant Area/Equipment Temperature Monitoring Program

    International Nuclear Information System (INIS)

    McGuire, L.L.

    1991-01-01

    The Perry Nuclear Power Plant Area/Equipment Temperature Monitoring Program serves two purposes. The first is to track temperature trends during normal plant operation in areas where suspected deviations from established environmental profiles exist. This includes the use of Resistance Temperature Detectors, Recorders, and Temperature Dots for evaluation of equipment qualified life for comparison with tested parameters and the established Environmental Design Profile. It also may be used to determine the location and duration of steam leaks for effect on equipment qualified life. The second purpose of this program is to aid HVAC design engineers in determining the source of heat outside anticipated design parameters. Resistance Temperature Detectors, Recorders, and Temperature Dots are also used for this application but the results may include design changes to eliminate the excess heat or provide qualified equipment (cable) to withstand the elevated temperature, splitting of environmental zones to capture accurate temperature parameters, or continued environmental monitoring for evaluation of equipment located in hot spots

  5. Temperature dependence of single-particle properties in nuclear matter

    International Nuclear Information System (INIS)

    Zuo, W.; Lu, G.C.; Li, Z.H.; Lombardo, U.; Schulze, H.-J.

    2006-01-01

    The single-nucleon potential in hot nuclear matter is investigated in the framework of the Brueckner theory by adopting the realistic Argonne V 18 or Nijmegen 93 two-body nucleon-nucleon interaction supplemented by a microscopic three-body force. The rearrangement contribution to the single-particle potential induced by the ground state correlations is calculated in terms of the hole-line expansion of the mass operator and provides a significant repulsive contribution in the low-momentum region around and below the Fermi surface. Increasing temperature leads to a reduction of the effect, while increasing density makes it become stronger. The three-body force suppresses somewhat the ground state correlations due to its strong short-range repulsion, increasing with density. Inclusion of the three-body force contribution results in a quite different temperature dependence of the single-particle potential at high enough densities as compared to that adopting the pure two-body force. The effects of three-body force and ground state correlations on the nucleon effective mass are also discussed

  6. 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

  7. High temperature continuous operation in the HTTR (HP-11). Summary of the test results in the high temperature operation mode

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Ueta, Shohei; Sumita, Junya; Goto, Minoru; Nakagawa, Shigeaki; Hamamoto, Shimpei; Tochio, Daisuke

    2010-11-01

    A high temperature (950 degrees C) continuous operation has been performed for 50 days on the HTTR from January to March in 2010, and the potential to supply stable heat of high temperature for hydrogen production for a long time was demonstrated for the first time in the world. JAEA has evaluated the experimental data obtained by this operation and past rated continuous one, and built the database necessary for commercial HTGRs. According to the results, the concentration of FP released from the fuels in the HTTR was a single through triple-digit lower than that in the foreign HTGRs. It became apparent that the fuels used in the HTTR are the best quality in the world. This successful operation could establish technological basis of HTGRs and show potential of nuclear energy as heat source for innovative thermo-chemical-based hydrogen production, emitting greenhouse gases on a 'low-carbon path' for the first time in the world. We have a plan to progress R and D for practical use of hydrogen production system with HTGRs in the future. (author)

  8. 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)

  9. Structural integrity and its role in nuclear safety recent UK developments in the development of high temperature design procedures

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1991-01-01

    The structural design rules for the reactors which operate at high temperature are not yet well developed. There is not difficulty in producing the plants which meet the high standards required by nuclear industry. However, there are the issues to be resolved which are associated with the deterioration of components in service, in order to achieve the optimum use of materials and the reduction of capital costs. The safety of plants is not at risk since any deterioration is detected by in-service monitoring, nevertheless, there would be severe economic penalty, if a plant must be retired prematurely because the continuing safety could not be demonstrated. In this paper, a liquid metal fast breeder reactor is taken up as an example, and the topics in which research plays a role for providing improved design rules are identified. Shakedown interaction diagrams, the methods of analysis based on shakedown, inelastic analysis and constitutive equations, creep fatigue damage and thermal shock, thermal striping, welds, defect assessment and so on are discussed. (K.I.)

  10. High Temperature Test Possibility at the HANARO Out-core Region through a Thermal Analysis

    International Nuclear Information System (INIS)

    Kang, Young-Hwan; Choi, Myung-Hwan; Cho, Man-Soon; Choo, Kee-Nam; Kim, Bong-Goo

    2007-01-01

    The development of an advanced reactor system such as a next generation nuclear plant and other generation IV systems require new fuels, claddings, and structural materials. To characterize the performance of these new materials, it is necessary for us to have a leading-edge technology to satisfy the specific test requirements such as the conditions of high neutron exposures and high operating temperatures. Thus, nuclear data on HANARO's vertical test holes have been gathered and reviewed to evaluate the usability of the test holes located at the out-core zone of HANARO. In 2007, neutron flux levels of the concerned test holes and the gamma heat of the specimens and two different specimen holder materials of Al and Mo at the concerned test hole were obtained to enhance the utilization of the HANARO reactor and to develop new design concepts for high temperature irradiation tests. Based on the data, a series of thermal analyses was implemented to provide a reasonable demonstration and guidance on limitations or application

  11. Radiation heat transfer within and from high temperature plumes composed of steam and molten nuclear debris

    International Nuclear Information System (INIS)

    Condiff, D.W.

    1987-03-01

    The Differential Approximation of Radiation Heat Transfer which includes anisotropic scattering is formulated to account for multiple source and temperature fields of multiphase flow. The formulation is applied to a simplified model of a plume consisting of high temperature emissive particles in steam at parametrically variable lower temperatures. Parametric model calculations are presented which account for spectral emission and absorption by steam using a band approximation as well as emission, absorption and scattering by the debris. The results are found to be far more sensitive to emission properties of individual particles, than to their scattering properties at high temperatures

  12. Effect of High Temperature or fire on heavy weight concrete properties used in nuclear facilities

    International Nuclear Information System (INIS)

    Sakr, K.

    2003-01-01

    In the present work the effect of different duration (1, 2 and 3 hours) of high temperatures (250 degree C, 500 degree C, 750 degree C and 950 degree C) on the physical and mechanical properties of heavy concrete shields were studied. The effect of fire fitting systems on ordinary concrete was investigated. The work was extended to determine the effect of high temperature or accidental fire on the radiation properties of heavy weight concrete. Results showed that ilmenite concrete had the highest density, absorption, and modulus of elasticity when compared to the other types of studied concrete and it had also higher values of compressive, tensile, bending and bonding strength than ordinary or baryte concrete. Ilmenite concrete had the highest attenuation of transmitted gamma rays in comparing to gravel concrete and baryte concrete. Ilmenite concrete was more resistant to elevated temperature than gravel concrete and baryte concrete. Foam or air as a fire fitting system in concrete structure that exposed to high temperature or accidental fire proved that better than water

  13. Estimation of Hydrazine Decomposition on Measuring the High-Temperature pH in Hydrazine/ETA Solutions at 553 K

    International Nuclear Information System (INIS)

    Hwang, Jae Sik; Yeon, Jei Won; Yun, Myung Hee; Song, Kyu Seok; Lee, Sang Ill

    2010-01-01

    Hydrazine is one of the most excellent oxygen scavengers used in the secondary circuit of nuclear power plants. Furthermore, in some pants, the hydrazine is used as a source of hydrogen required to suppress radiolysis of the coolant water in the primary loop. When hydrazine was exposed in the high temperature and high pressure water, it can be decomposed into the various products such as NH 3 , N 2 , H 2 , and NO 3 ions. As the result, the pH of solution containing hydrazine in the condition of the high temperature and high pressure can be changed by those decomposed products. In the present work, we investigated the decomposition behavior of hydrazine in ETA (ethanol amine) solution. In addition, we measured the high temperature pH at 553 K on the various hydrazine/ETA solutions for confirming the applicability of the yttria stabilized zirconia (YSZ)- based pH electrode in secondary circuit of the nuclear power plants

  14. 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.)

  15. Safety of nuclear reactors - Part A - unsteady state temperature history mathematical model

    International Nuclear Information System (INIS)

    El-Shayeb, M.; Yusoff, M.Z.; Boosroh, M.H.; Ideris, F.; Hasmady Abu Hassan, S.; Bondok, A.

    2004-01-01

    A nuclear reactor structure under abnormal operations of near meltdown will be exposed to a tremendous amount of heat flux in addition to the stress field applied under normal operation. Temperature encountered in such case is assumed to be beyond 1000 Celsius degrees. A 2-dimensional mathematical model based on finite difference methods, has been developed for the fire resistance calculation of a concrete-filled square steel column with respect to its temperature history. Effects due to nuclear radiation and mechanical vibrations will be explored in a later future model. The temperature rise in each element can be derived from its heat balance by applying the parabolic unsteady state, partial differential equation and numerical solution into the steel region. Calculation of the temperature of the elementary regions needs to satisfy the symmetry conditions and the relevant material properties. The developed mathematical model is capable to predict the temperature history in the column and on the surface with respect to time. (authors)

  16. Damage-Tolerant, Lightweight, High-Temperature Radiator for Nuclear Powered Spacecraft

    Data.gov (United States)

    National Aeronautics and Space Administration — Game-changing propulsion systems are often enabled by novel designs using advanced materials. Radiator performance dictates power output for nuclear electric...

  17. 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

  18. Two decades on[Research into high-temperature superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Durrani, M. [Physics World (United Kingdom)

    2006-04-15

    Research into high-temperature superconductors should focus on experiment, not theory. While the world looked on in horror at the events unfolding at the Chernobyl nuclear-power plant in the Soviet Union 20 years ago this month, another significant - but far less reported - development in the world of physics had just taken place. On 17 April 1986 a short paper by Georg Bednorz and Alexander Mueller arrived at the offices of Zeitschrift fuer Physik in Heidelberg, Germany. The two physicists, based at IBM's Zurich Research Laboratory in Switzerland, announced they had made a material from barium, lanthanum, copper and oxygen that could conduct electricity without resistance when cooled below a transition temperature, T{sub c}, of about 30 K. It was the world's first 'high-temperature' superconductor. Driven by the dream of materials that can superconduct at room temperature, experimentalists scurried back to their labs. Within a year, a T{sub c} of 90 K in another material had been reported and by October 1987 Bednorz and Mueller had been crowned with a Nobel prize. While papers on high-temperature superconductivity have continued to stream out since those heady days, progress has been slower than expected. Applications like levitating trains and resistance-free power cables are only now starting to come to market. Scientists have been unable to make superconducting wires that work much above 130 K, while a reliable theory of high-temperature superconductivity remains elusive. Even if we had such a theory, it is not clear that it would predict which materials might superconduct at room temperature. After all, the Bardeen-Cooper-Schrieffer theory, which explains the behaviour of low-temperature superconductors with admirable success, said nothing about the superconducting properties of Bednorz and Mueller's copper-oxide ceramics. What successes there have been over the last 20 years - such as the recent discoveries that iron, single crystals

  19. Temperature measurements inside nuclear reactor cores

    International Nuclear Information System (INIS)

    Tarassenko, Serge

    1969-11-01

    Non negligible errors may happen in nuclear reactor temperature measurements using magnesium oxide insulated and stainless steel sheathed micro-wire thermocouples, when these thermometric lines are placed under operational conditions typical of electrical power stations. The present work shows that this error is principally due to electrical hysteresis and polarization phenomena in the insulator subjected to the strong fields generated by common-mode voltages. These phenomena favour the unsymmetrical common-mode current flow and thus lead to the differential-mode voltage generation which is superposing on the thermoelectric hot junction potential. A calculation and an experimental approach make possible the importance of the magnesium oxide insulating characteristics, the hot junction insulation, the choice of the main circuits in the data processing equipment as well as the galvanic isolation performances and the common-mode rejection features of all the measurement circuits. A justification is thereby given for the severe conditions imposed for the acceptance of thermoelectric materials; some particular precautions to be taken are described, as well as the high performance characteristics which have to be taken into account in choosing measurement systems linked to thermometric circuits with sheathed micro-wire thermocouples. (author) [fr

  20. A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Beach, N. J.; Knapp, S. M. M.; Landis, C. R., E-mail: landis@chem.wisc.edu [Department of Chemistry, University of Wisconsin-Madison, Madison, Wisconsin 53719 (United States)

    2015-10-15

    The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from −90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor to be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.

  1. 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.

  2. 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.

  3. Lifetime improvement of sheathed thermocouples for use in high-temperature and thermal transient operations

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Clift, J.H.

    1982-01-01

    Premature failure of small-diameter, magnesium-oxide-insulated sheathed thermocouples occurred when they were placed within nuclear fuel rod simulators (FRSs) to measure high temperatures and to follow severe thermal transients encountered during simulation of nuclear reactor accidents in Oak Ridge National Laboratory (ORNL) thermal-hydraulic test facilities. Investigation of thermally cycled thermocouples yielded three criteria for improvement of thermocouple lifetime: (1) reduction of oxygen impurities prior to and during their fabrication, (2) refinement of thermoelement grain size during their fabrication, and (3) elimination of prestrain prior to use above their recrystallization temperature. The first and third criteria were satisfied by improved techniques of thermocouple assembly and by a recovery anneal prior to thermocouple use

  4. Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de.

    1986-01-01

    The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied, applied to a nine heated tube bundle experimental data set. (Author) [pt

  5. Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de.

    1986-01-01

    The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied to a nine heated tube bundle experimental data set. (Author) [pt

  6. Crystal structure and ionic conduction path of solid electrolytic materials by high temperature neutron diffraction method

    International Nuclear Information System (INIS)

    Yashima, Masatomo; Nomura, Katsuhiro

    2005-01-01

    Research of the distribution of oxide ions and the ionic conduction path of bismuth oxide (Bi 2 O 3 ), cerium oxide (CeO 2 ) and lanthanum gallate ((La 0.8 Sr 0.2 )(Ga 0.8 Mg 0.15 Co 0.05 )O 3-δ ) is stated. The high temperature neutron diffraction method, analytical method such as Rietveld method, crystal structure analysis of ionic conductor and MEM (Maximum- Entropy Method) are explained. The nuclear density distribution of oxide ions in bismuth oxide showed so larger distribution in the direction of and than Bi ions that the oxide ions conducted these direction in the crystal. The nuclear density distribution of oxide ions of cerium oxide indicated larger distribution in the direction of than Ce ions and its tendency was remarkable at high temperature. Accordingly, the oxide ions conducted in the direction of and . The oxide ions distribution in lanthanum gallate compound was larger and complicated than positive ions. The oxide ions conducted to by describing an arc between the two stable positions. The nuclear density on the conduction path increased with increasing temperature. This above result corresponded to increase of oxide ion conductivity in the area. (S.Y.)

  7. Preliminary estimations on the heat recovery method for hydrogen production by the high temperature steam electrolysis

    International Nuclear Information System (INIS)

    Koh, Jae Hwa; Yoon, Duck Joo

    2009-01-01

    As a part of the project 'development of hydrogen production technologies by high temperature electrolysis using very high temperature reactor', we have developed an electrolyzer model for high temperature steam electrolysis (HTSE) system and carried out some preliminary estimations on the effects of heat recovery on the HTSE hydrogen production system. To produce massive hydrogen by using nuclear energy, the HTSE process is one of the promising technologies with sulfur-iodine and hybrid sulfur process. The HTSE produces hydrogen through electrochemical reaction within the solid oxide electrolysis cell (SOEC), which is a reverse reaction of solid oxide fuel cell (SOFC). The HTSE system generally operates in the temperature range of 700∼900 .deg. C. Advantages of HTSE hydrogen production are (a) clean hydrogen production from water without carbon oxide emission, (b) synergy effect due to using the current SOFC technology and (c) higher thermal efficiency of system when it is coupled nuclear reactor. Since the HTSE system operates over 700 .deg. C, the use of heat recovery is an important consideration for higher efficiency. In this paper, four different heat recovery configurations for the HTSE system have been investigated and estimated

  8. Electrochemical applications of room temperature ionic liquids in nuclear fuel cycle

    International Nuclear Information System (INIS)

    Venkatesan, K.A.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2008-01-01

    Applications of room temperature ionic liquids (RTILs) have invaded all branches of science. They are also receiving an upsurge, in recent years, for possible applications in various stages of nuclear fuel cycle. Ionic liquids are compounds composed entirely of ions existing in liquid state and RTILs are ionic liquids molten at temperatures lower than 373 K. RTILs are generally made up of an organic cation and an inorganic or an organic anion. Room temperature ionic liquids have several fascinating properties, which are unique to a particular combination of cation and anion. The properties such as insignificant vapor pressure, amazing ability to dissolve organic and inorganic compounds, wide electrochemical window are the specific advantages when dealing with application of RTILs for reprocessing of spent nuclear fuel. The ionic liquids are regarded as designer or tailor-made solvents as their properties can be tuned for desired application by appropriate cation-anion combinations. An excellent review by Wilkes describes about the historical perspectives of room temperature ionic liquids, pioneers in that area, events and the products delivered till 2001. Furthermore, several comprehensive reviews have been made on room temperature ionic liquids by various authors

  9. Small reactors for low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    1988-06-01

    In accordance with the Member States' calls for information exchange in the field of nuclear heat application (NHA) two IAEA meetings were organized already in 1976 and 1977. After this ''promising period'', the development of relevant programmes in IAEA Member States was slowed down and therefore only after several years interruption a new Technical Committee Meeting with a Workshop was organized in late 1983, to review the status of NHA, after a few new specific plans appeared in some IAEA Member States in the early 1980's for the use of heat from existing or constructed NPPs and for developing nuclear heating plants (NHP). In June 1987 an Advisory Group Meeting was convened in Winnipeg, Canada, to discuss and formulate a state-of-the-art review on ''Small Reactors for Low Temperature Nuclear Heat Application''. Information on this subject gained up to 1987 in the Member States whose experts attended this meeting is embodied in the present Technical Report. Figs and tabs

  10. 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)

  11. Advances in high temperature chemistry

    CERN Document Server

    Eyring, Leroy

    1969-01-01

    Advances in High Temperature Chemistry, Volume 2 covers the advances in the knowledge of the high temperature behavior of materials and the complex and unfamiliar characteristics of matter at high temperature. The book discusses the dissociation energies and free energy functions of gaseous monoxides; the matrix-isolation technique applied to high temperature molecules; and the main features, the techniques for the production, detection, and diagnosis, and the applications of molecular beams in high temperatures. The text also describes the chemical research in streaming thermal plasmas, as w

  12. 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)

  13. Prospects for sub-micron solid state nuclear magnetic resonance imaging with low-temperature dynamic nuclear polarization.

    Science.gov (United States)

    Thurber, Kent R; Tycko, Robert

    2010-06-14

    We evaluate the feasibility of (1)H nuclear magnetic resonance (NMR) imaging with sub-micron voxel dimensions using a combination of low temperatures and dynamic nuclear polarization (DNP). Experiments are performed on nitroxide-doped glycerol-water at 9.4 T and temperatures below 40 K, using a 30 mW tunable microwave source for DNP. With DNP at 7 K, a 0.5 microL sample yields a (1)H NMR signal-to-noise ratio of 770 in two scans with pulsed spin-lock detection and after 80 db signal attenuation. With reasonable extrapolations, we infer that (1)H NMR signals from 1 microm(3) voxel volumes should be readily detectable, and voxels as small as 0.03 microm(3) may eventually be detectable. Through homonuclear decoupling with a frequency-switched Lee-Goldburg spin echo technique, we obtain 830 Hz (1)H NMR linewidths at low temperatures, implying that pulsed field gradients equal to 0.4 G/d or less would be required during spatial encoding dimensions of an imaging sequence, where d is the resolution in each dimension.

  14. 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)

  15. High-temperature superconductivity

    International Nuclear Information System (INIS)

    Lynn, J.W.

    1990-01-01

    This book discusses development in oxide materials with high superconducting transition temperature. Systems with Tc well above liquid nitrogen temperature are already a reality and higher Tc's are anticipated. The author discusses how the idea of a room-temperature superconductor appears to be a distinctly possible outcome of materials research

  16. Searching for superconductors with high critical temperature

    Energy Technology Data Exchange (ETDEWEB)

    Chao, C

    1977-08-18

    Critical temperature of superconductors can be and must be raised so that their range of application can be broadened. It was estimated that, in 3 to 5 years, superconductor electric generators might be used in nuclear submarines and/or other applications where the requirements of small volume and light weight are critical. The BCS theory was recapitulated. Possible methods of achieving higher critical temperature were proposed and discussed.

  17. Symmetric and asymmetric nuclear matter in the Thomas-Fermi model at finite temperatures

    International Nuclear Information System (INIS)

    Strobel, K.; Weber, F.; Weigel, M.K.

    1999-01-01

    The properties of warm symmetric and asymmetric nuclear matter are investigated in the frame of the Thomas-Fermi approximation using a recent modern parameterization of the effective nucleon-nucleon interaction of Myers and Swiatecki. Special attention is paid to the liquid-gas phase transition, which is of special interest in modern nuclear physics. We have determined the critical temperature, critical density and the so-called flash temperature. Furthermore, the equation of state for cold neutron star matter is calculated. (orig.)

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. High temperature ductility of austenitic alloys exposed to thermal neutrons

    International Nuclear Information System (INIS)

    Watanabe, K.; Kondo, T.; Ogawa, Y.

    1982-01-01

    Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

  3. Some comments on the behaviour of the excited nuclear matter formed in nuclear collisions at high energies

    International Nuclear Information System (INIS)

    Besliu, Calin; Jipa, Alexandru; Argintaru, Dan

    2003-01-01

    In the last years many experiments have been performed in different laboratories to investigate the behaviour of the nuclear matter formed in nuclear collisions at high energies. Therefore, many experimental results are available at present. For explaining these experimental results a lot of models have been proposed. A very large number of concepts have been used. Taking into account some own experimental results obtained in proton-nucleus and nucleus-nucleus collisions at energies between a few A GeV and a few hundred A GeV we comments in the frame a phenomenological geometric picture the main experimental results on charged particle multiplicities, participants, cross sections, momentum spectra, temperature slopes, nuclear matter flow, size and structure of the participant regions, antiparticle to particle ratios and chemical potential. Some jumps in the dependencies of some interesting physical quantities on the available energies in the centre of mass system can be reported. Trends to behaviours like-saturation of some physical quantities are observed, too. Therefore, some connections with the possible phase transitions in nuclear matter are included. A few specific signals of different phase transitions in nuclear matter are suggested. (authors)

  4. Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

    Science.gov (United States)

    Andre, Matthieu A.; Bardet, Philippe M.; Burns, Ross A.; Danehy, Paul M.

    2015-01-01

    The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

  5. High Temperature Transducers for Online Monitoring of Microstructure Evolution

    Energy Technology Data Exchange (ETDEWEB)

    Lissenden, Cliff [Pennsylvania State Univ., State College, PA (United States); Tittmann, Bernhard [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

    2015-03-30

    A critical technology gap exists relative to online condition monitoring (CM) of advanced nuclear plant components for damage accumulation; there are not capable sensors and infrastructure available for the high temperature environment. The sensory system, monitoring methodology, data acquisition, and damage characterization algorithm that comprise a CM system are investigated here. Thus this work supports the DOE mission to develop a fundamental understanding of advanced sensors to improve physical measurement accuracy and reduce uncertainty. The research involves a concept viability assessment, a detailed technology gap analysis, and a technology development roadmap.

  6. High temperature ultrasonic sensor for fission gas characterization in MTR harsh environment

    Science.gov (United States)

    Gatsa, O.; Combette, P.; Rozenkrantz, E.; Fourmentel, D.; Destouches, C.; Ferrandis, J. Y. AD(; )

    2018-01-01

    In the contemporary world, the measurements in hostile environment is one of the predominant necessity for automotive, aerospace, metallurgy and nuclear plant. The measurement of different parameters in experimental reactors is an important point in nuclear power strategy. In the near past, IES (Institut d'Électronique et des Systèmes) on collaboration with CEA (Commissariat à l'Energie Atomique et aux Energies Alternatives) have developed the first ultrasonic sensor for the application of gas quantity determination that has been tested in a Materials Testing Reactor (MTR). Modern requirements state to labor with the materials that possess stability on its parameters around 350°C in operation temperature. Previous work on PZT components elaboration by screen printing method established the new basis in thick film fabrication and characterization in our laboratory. Our trials on Bismuth Titanate ceramics showed the difficulties related to high electrical conductivity of fabricated samples that postponed further research on this material. Among piezoceramics, the requirements on finding an alternative solution on ceramics that might be easily polarized and fabricated by screen printing approach were resolved by the fabrication of thick film from Sodium Bismuth Titanate (NBT) piezoelectric powder. This material exhibits high Curie temperature, relatively good piezoelectric and coupling coefficients, and it stands to be a good solution for the anticipated application. In this paper, we present NBT thick film fabrication by screen printing, characterization of piezoelectric, dielectric properties and material parameters studies in dependence of temperature. Relatively high resistivity in the range of 1.1013 Ohm.cm for fabricated thick film is explained by Aurivillius structure in which a-and b-layers form perovskite structure between oxides of c-layer. Main results of this study are presented and discussed in terms of feasibility for an application to a new sensor

  7. 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

  8. 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

  9. High Total Ionizing Dose and Temperature Effects on Micro- and Nano-electronic Devices

    International Nuclear Information System (INIS)

    Gaillardin, M.; Martinez, M.; Paillet, P.; Leray, J.L.; Marcandella, C.; Duhamel, O.; Raine, M.; Richard, N.; Girard, S.; Ouerdane, Y.; Boukenter, A.; Goiffon, V.; Magnan, P.; Andrieu, F.; Barraud, S.; Faynot, O.

    2013-06-01

    This paper investigates the vulnerability of several micro- and nano-electronic technologies to a mixed harsh environment including high total ionizing dose at MGy levels and high temperature. Such operating conditions have been revealed recently for several applications like new security systems in existing or future nuclear power plants, fusion experiments, or deep space missions. In this work, the competing effects already reported in literature of ionizing radiations and temperature are characterized in elementary devices made of MOS transistors from several technologies. First, devices are irradiated using a radiation laboratory X-ray source up to MGy dose levels at room temperature. Devices are grounded during irradiation to simulate a circuit which waits for a wake up signal, representing most of the lifetime of an integrated circuit operating in a harsh environment. Devices are then annealed at several temperatures to discuss the post-irradiation behavior and to determine whether an elevated temperature is an issue or not for circuit function in mixed harsh environments. (authors)

  10. Estimation of characteristics on high temperature filtration system for particle removal in vitrification process

    International Nuclear Information System (INIS)

    Park, Seung Chul; Ryu, Bo Hyun; Park, Byoung Chul; Ryu, Chang Soo; Hwang, Tae Won; Ha, Jong Hyun

    2003-01-01

    High temperature filtration technology has been widely used in nuclear industry systems to remove particulate matter from air and gas streams. Air filters are defined as porous structures through which air is passed to separate out entrained particulate matter. Especially among of them, ceramic candle filters are suitable to gain efficient dust removal at high temperatures and achieve high collection efficiencies for (sub-)micron particles. The paper presents experimental results for their application in the pilot scale vitrification plant operations. Experimental results were transformed into design equations for (i) total pressure drop and the effect of face velocity; (ii) the prediction of the operating parameters

  11. Temperature conditions of foundation plates under nuclear power plant reactor compartments

    International Nuclear Information System (INIS)

    Ehsaulov, S.L.

    1990-01-01

    Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

  12. 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

  13. Proceedings of the 8. National Seminar on Technology and Safety of Nuclear Power Plants and Nuclear Facilities

    International Nuclear Information System (INIS)

    Antariksawan, Anhar R.; Soetrisnanto, Arnold Y.; Aziz, Ferhat; Untoro, Pudji; Su'ud, Zaki; Zarkasi, Amin Santosa; Umar, Faraz H.; Teguh Bambang; Hafnan, M.; Mustafa, Bustani; Rosfian, H.

    2002-10-01

    The eight proceeding of National Seminar on Technology and Safety of Nuclear Power Plant and Nuclear Facilities held by National Atomic Energy Agency and University of Trisakti. The aims of Seminar is to exchange and disseminate information about safety and nuclear Power Plant Temperature Reactor and Application for National Development sustain able and High Technology. This Seminar covers all aspect Technology, Power Reactor : Research Reactor; High Temperature Reactor and Nuclear Facilities. There are 33 articles have separated index

  14. Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

  15. Summary report on technical experiences from high-temperature helium turbomachinery testing in Germany

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1996-01-01

    In Germany a comprehensive research and development program was initiated in 1968 for a Brayton (closed) cycle power conversion system. The program was for ultimate use with a high temperature, helium cooled nuclear reactor heat source (the HHT project) for electricity generation using helium as the working fluid. The program continued until 1982 in international cooperation with the United States and Switzerland. This document describes the designs and reports the results of testing activities that addressed the development of turbines, compressors, hot gas ducts, materials, heat exchangers and other equipment items for use with a helium working fluid at high temperatures. 67 refs, 34 figs, tabs

  16. 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

  17. Research of management information system of radiation protection for low temperature nuclear heating reactor

    International Nuclear Information System (INIS)

    Bai Hongtao; Wang Jiaying; Wu Manxue

    2001-01-01

    Management information system of radiation protection for low temperature reactor uses computer to manage the data of the low temperature nuclear heating reactor radiation monitoring, it saves the data from the front real-time radiation monitoring system, comparing these data with historical data to give the consequence. Also, the system provides some picture in order to show space information at need. The system, based on Microsoft Access 97, consists of nine parts, including radiation dose, environmental data, meteorological data and so on. The system will have value in safely operation of the low temperature nuclear heating reactor

  18. Evolution of elevated containment temperatures at Calvert Cliffs Nuclear Power Plant

    International Nuclear Information System (INIS)

    Branch, R.D. Jr.

    1991-01-01

    In this paper the author describes the events which caused Calvert Cliffs Nuclear Power Plant engineers to recognize a need for monitoring of ambient temperatures within containment. The early attempts at temperature monitoring programs are discussed and critiqued. Primary failings of these early programs included a failure to collect temperature data under a variety of external conditions and a lack of quality assurance to make the data useful for design change. From these early attempts Calvert Cliffs developed a new, extensive temperature monitoring program designed to collect data over a two-year period. The author outlines the planned temperature monitoring program and discusses its expected results

  19. 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)

  20. 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)

  1. Nuclear energy in metallurgy

    Energy Technology Data Exchange (ETDEWEB)

    Jirak, Z; Malik, J; Vrba, J

    1976-01-01

    The present power situation and its estimated development with a view to metallurgy is presented. The possibilities of the development of Czechoslovak metallurgy are described with regard to conventional fuels and to nuclear power applications. The programme of the use of nuclear power in countries with a highly developed metallurgical industry, such as Japan, the FRG, etc., is presented and the technical pre-requisites for the use of nuclear power in metallurgy, namely the use of high temperature reactors and their incorporation in nuclear metallurgical complexes are discussed. The problems are indicated of the selection of suitable materials for high temperature reactors and the experience is described with the operation of such equipment. The results are given of the analysis of 10 variants of the model of a nuclear metallurgical complex manufacturing 1000 tons of sponge iron per day and having four main technological circuits (the helium circuit, the steam circuit, the reduction gas circuit and the cycle of metallurgical processes). An estimate is given of the capital costs of building a high temperature reactor, a power plant and a metallurgical complex with the reactor. The costs are also given of steel and power production in a nuclear metallurgical complex.

  2. Probabilistic residual life assessment of high temperature pipings in nuclear power plants against creep fatigue damage: final report

    International Nuclear Information System (INIS)

    Gupta, Sanjay

    2014-02-01

    Residual life assessment of components of nuclear power plants is essential for their operational safety, reliability and financial viability. The high risks involved in the event of failures in nuclear power plants have led to the development of design philosophies that incorporate extreme conservatism in design. The implications of such conservatism in design leads to more frequent maintenance operations than necessary

  3. Conversion of nuclear waste to molten glass: Formation of porous amorphous alumina in a high-Al melter feed

    Science.gov (United States)

    Xu, Kai; Hrma, Pavel; Washton, Nancy; Schweiger, Michael J.; Kruger, Albert A.

    2017-01-01

    The transition of Al phases in a simulated high-Al high-level nuclear waste melter feed heated at 5 K min-1 to 700 °C was investigated with transmission electron microscopy, 27Al nuclear magnetic resonance spectroscopy, the Brunauer-Emmett-Teller method, and X-ray diffraction. At temperatures between 300 and 500 °C, porous amorphous alumina formed from the dehydration of gibbsite, resulting in increased specific surface area of the feed (∼8 m2 g-1). The high-surface-area amorphous alumina formed in this manner could potentially stop salt migration in the cold cap during nuclear waste vitrification.

  4. An investigation on high temperature fatigue properties of tempered nuclear-grade deposited weld metals

    Science.gov (United States)

    Cao, X. Y.; Zhu, P.; Yong, Q.; Liu, T. G.; Lu, Y. H.; Zhao, J. C.; Jiang, Y.; Shoji, T.

    2018-02-01

    Effect of tempering on low cycle fatigue (LCF) behaviors of nuclear-grade deposited weld metal was investigated, and The LCF tests were performed at 350 °C with strain amplitudes ranging from 0.2% to 0.6%. The results showed that at a low strain amplitude, deposited weld metal tempered for 1 h had a high fatigue resistance due to high yield strength, while at a high strain amplitude, the one tempered for 24 h had a superior fatigue resistance due to high ductility. Deposited weld metal tempered for 1 h exhibited cyclic hardening at the tested strain amplitudes. Deposited weld metal tempered for 24 h exhibited cyclic hardening at a low strain amplitude but cyclic softening at a high strain amplitude. Existence and decomposition of martensite-austenite (M-A) islands as well as dislocations activities contributed to fatigue property discrepancy among the two tempered deposited weld metal.

  5. FEM Modeling of the Relationship between the High-Temperature Hardness and High-Temperature, Quasi-Static Compression Experiment.

    Science.gov (United States)

    Zhang, Tao; Jiang, Feng; Yan, Lan; Xu, Xipeng

    2017-12-26

    The high-temperature hardness test has a wide range of applications, but lacks test standards. The purpose of this study is to develop a finite element method (FEM) model of the relationship between the high-temperature hardness and high-temperature, quasi-static compression experiment, which is a mature test technology with test standards. A high-temperature, quasi-static compression test and a high-temperature hardness test were carried out. The relationship between the high-temperature, quasi-static compression test results and the high-temperature hardness test results was built by the development of a high-temperature indentation finite element (FE) simulation. The simulated and experimental results of high-temperature hardness have been compared, verifying the accuracy of the high-temperature indentation FE simulation.The simulated results show that the high temperature hardness basically does not change with the change of load when the pile-up of material during indentation is ignored. The simulated and experimental results show that the decrease in hardness and thermal softening are consistent. The strain and stress of indentation were analyzed from the simulated contour. It was found that the strain increases with the increase of the test temperature, and the stress decreases with the increase of the test temperature.

  6. FEM Modeling of the Relationship between the High-Temperature Hardness and High-Temperature, Quasi-Static Compression Experiment

    Directory of Open Access Journals (Sweden)

    Tao Zhang

    2017-12-01

    Full Text Available The high-temperature hardness test has a wide range of applications, but lacks test standards. The purpose of this study is to develop a finite element method (FEM model of the relationship between the high-temperature hardness and high-temperature, quasi-static compression experiment, which is a mature test technology with test standards. A high-temperature, quasi-static compression test and a high-temperature hardness test were carried out. The relationship between the high-temperature, quasi-static compression test results and the high-temperature hardness test results was built by the development of a high-temperature indentation finite element (FE simulation. The simulated and experimental results of high-temperature hardness have been compared, verifying the accuracy of the high-temperature indentation FE simulation.The simulated results show that the high temperature hardness basically does not change with the change of load when the pile-up of material during indentation is ignored. The simulated and experimental results show that the decrease in hardness and thermal softening are consistent. The strain and stress of indentation were analyzed from the simulated contour. It was found that the strain increases with the increase of the test temperature, and the stress decreases with the increase of the test temperature.

  7. 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)

  8. Effects of nuclear forces on ion thermalization in high-temperature plasmas

    International Nuclear Information System (INIS)

    Gould, R.J.

    1982-01-01

    The energy loss rate is computed for a fast nonrelativistic ion traversing a plasma, with special emphasis on the basic problem of proton-proton ( p-p) interactions. Elastic p-p scattering is described in terms of the effective range r 0 and scattering length a 0 associated with the nuclear interaction. The nuclear s-wave phase shift (delta 0 ) in the presence of the Coulomb field is computed as a function of energy E 0 from the r 0 -a 0 formalism; delta 0 has a very weak energy dependence, essentially logarithmic, in the 1-100 MeV domain

  9. 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…

  10. Supersymmetry at high temperatures

    International Nuclear Information System (INIS)

    Das, A.; Kaku, M.

    1978-01-01

    We investigate the properties of Green's functions in a spontaneously broken supersymmetric model at high temperatures. We show that, even at high temperatures, we do not get restoration of supersymmetry, at least in the one-loop approximation

  11. Artificial neural networks in prediction of mechanical behavior of concrete at high temperature

    International Nuclear Information System (INIS)

    Mukherjee, A.; Nag Biswas, S.

    1997-01-01

    The behavior of concrete structures that are exposed to extreme thermo-mechanical loading is an issue of great importance in nuclear engineering. The mechanical behavior of concrete at high temperature is non-linear. The properties that regulate its response are highly temperature dependent and extremely complex. In addition, the constituent materials, e.g. aggregates, influence the response significantly. Attempts have been made to trace the stress-strain curve through mathematical models and rheological models. However, it has been difficult to include all the contributing factors in the mathematical model. This paper examines a new programming paradigm, artificial neural networks, for the problem. Implementing a feedforward network and backpropagation algorithm the stress-strain relationship of the material is captured. The neural networks for the prediction of uniaxial behavior of concrete at high temperature has been presented here. The results of the present investigation are very encouraging. (orig.)

  12. Design of the steam reformer for the HTR-10 high temperature process heat application

    International Nuclear Information System (INIS)

    Ju Huaiming; Xu Yuanhui; Jia Haijun

    2000-01-01

    The 10 MW High Temperature Reactor Test Module (HTR-10) is being constructed now and planned to be operational in 2000. One of the objectives is to develop the high temperature process heat application. The methane steam reformer is one of the key-facilities for the nuclear process heat application system. The paper describes the conceptual design of the HTR-10 Steam Reformer with He heating, and the design optimization computer code. It can be used to perform sensitivity analysis for parameters, and to improve the design. Principal parameters and construction features of the HTR-10 reformer heated by He are introduced. (author)

  13. 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)

  14. DC modeling and characterization of AlGaAs/GaAs heterojunction bipolar transistors for high-temperature applications

    International Nuclear Information System (INIS)

    Dikmen, C.T.; Dogan, N.S.; Osman, M.A.

    1994-01-01

    There is currently a demand for active electronic devices operating reliably over wide range of temperatures. Potential applications for the high-temperature devices and integrated circuits are in the areas of jet engine and control instrumentation for nuclear power plants. Here, the large signal dc characteristics of AlGaAs/GaAs heterojunction bipolar transistors (HBT) at high temperatures (27--300 C) are reported. A high-temperature SPICE model is developed which includes the recombination-generation current components and avalanche multiplication which become extremely important at high temperatures. The effect of avalanche breakdown is also included to model the current due to thermal generation of electron/hole pairs causing breakdown at high temperatures. A parameter extraction program is developed used to extract the model parameters of HBT's at different temperatures. Fitting functions for the model parameters as a function of temperature are developed. These parameters are then used in the SPICE Ebers-Moll model for the dc characterization of the HBT at any temperature between (27--300 C)

  15. Nuclear matter at high density: Phase transitions, multiquark states, and supernova outbursts

    International Nuclear Information System (INIS)

    Krivoruchenko, M. I.; Nadyozhin, D. K.; Rasinkova, T. L.; Simonov, Yu. A.; Trusov, M. A.; Yudin, A. V.

    2011-01-01

    Phase transition from hadronic matter to quark-gluon matter is discussed for various regimes of temperature and baryon number density. For small and medium densities, the phase transition is accurately described in the framework of the Field Correlation Method, whereas at high density predictions are less certain and leave room for the phenomenological models. We study formation of multiquark states (MQS) at zero temperature and high density. Relevant MQS components of the nuclear matter can be described using a previously developed formalism of the quark compound bags (QCB). Partialwave analysis of nucleon-nucleon scattering indicates the existence of 6QS which manifest themselves as poles of P matrix. In the framework of the QCB model, we formulate a self-consistent system of coupled equations for the nucleon and 6QS propagators in nuclear matter and the G matrix. The approach provides a link between high-density nuclear matter with the MQS components and the cumulative effect observed in reactions on the nuclei, which requires the admixture of MQS in the wave functions of nuclei kinematically. 6QS determines the natural scale of the density for a possible phase transition into theMQS phase of nuclear matter. Such a phase transition can lead to dynamic instability of newly born protoneutron stars and dramatically affect the dynamics of supernovae. Numerical simulations show that the phase transition may be a good remedy for the triggering supernova explosions in the spherically symmetric supernovamodels. A specific signature of the phase transition is an additional neutrino peak in the neutrino light curve. For a Galactic core-collapse supernova, such a peak could be resolved by the present neutrino detectors. The possibility of extracting the parameters of the phase of transition from observation of the neutrino signal is discussed also.

  16. Nuclear matter at high density: Phase transitions, multiquark states, and supernova outbursts

    Energy Technology Data Exchange (ETDEWEB)

    Krivoruchenko, M. I.; Nadyozhin, D. K.; Rasinkova, T. L.; Simonov, Yu. A.; Trusov, M. A., E-mail: trusov@itep.ru; Yudin, A. V. [Institute for Theoretical and Experimental Physics (Russian Federation)

    2011-03-15

    Phase transition from hadronic matter to quark-gluon matter is discussed for various regimes of temperature and baryon number density. For small and medium densities, the phase transition is accurately described in the framework of the Field Correlation Method, whereas at high density predictions are less certain and leave room for the phenomenological models. We study formation of multiquark states (MQS) at zero temperature and high density. Relevant MQS components of the nuclear matter can be described using a previously developed formalism of the quark compound bags (QCB). Partialwave analysis of nucleon-nucleon scattering indicates the existence of 6QS which manifest themselves as poles of P matrix. In the framework of the QCB model, we formulate a self-consistent system of coupled equations for the nucleon and 6QS propagators in nuclear matter and the G matrix. The approach provides a link between high-density nuclear matter with the MQS components and the cumulative effect observed in reactions on the nuclei, which requires the admixture of MQS in the wave functions of nuclei kinematically. 6QS determines the natural scale of the density for a possible phase transition into theMQS phase of nuclear matter. Such a phase transition can lead to dynamic instability of newly born protoneutron stars and dramatically affect the dynamics of supernovae. Numerical simulations show that the phase transition may be a good remedy for the triggering supernova explosions in the spherically symmetric supernovamodels. A specific signature of the phase transition is an additional neutrino peak in the neutrino light curve. For a Galactic core-collapse supernova, such a peak could be resolved by the present neutrino detectors. The possibility of extracting the parameters of the phase of transition from observation of the neutrino signal is discussed also.

  17. High temperature high vacuum creep testing facilities

    International Nuclear Information System (INIS)

    Matta, M.K.

    1985-01-01

    Creep is the term used to describe time-dependent plastic flow of metals under conditions of constant load or stress at constant high temperature. Creep has an important considerations for materials operating under stresses at high temperatures for long time such as cladding materials, pressure vessels, steam turbines, boilers,...etc. These two creep machines measures the creep of materials and alloys at high temperature under high vacuum at constant stress. By the two chart recorders attached to the system one could register time and temperature versus strain during the test . This report consists of three chapters, chapter I is the introduction, chapter II is the technical description of the creep machines while chapter III discuss some experimental data on the creep behaviour. Of helium implanted stainless steel. 13 fig., 3 tab

  18. Verification of the high temperature phase by the electron pair measurement at RHIC

    International Nuclear Information System (INIS)

    Akiba, Yasuyuki

    2013-01-01

    At the high energy nuclear collisions of the RHIC accelerator, the high density parton materials are created. If the matter is the quark gluon plasma (QGP) in the high temperature phase of the QCD, thermal photons are expected to be to be radiated there. The direct photon production from the gold + gold collision reactions at RHIC has been measured by using the 'virtual photon method'. In the gold + gold collisions, very many photons are produced compared with the p + p collisions. The production of the excess direct photons approximately agrees with the theoretical prediction of the thermal photon production from the initial temperature from 300 to 600 MeV QGP. In the present explanatory text, the direct photon measurements at the RHENIX experiments of RHIC are described starting from the discovery of high density matter by RHIC. The photon measurements which give direct evidence of the high temperature state and the virtual photon measurement method are reported briefly. The measurements of the direct photons and the estimation of the initial temperature at RHIC are described in detail with illustrations. Finally, some recent results are added and the ALICE experiments of LHC are referred. (S. Funahashi)

  19. Hydrogen production through high-temperature electrolysis in a solid oxide cell

    International Nuclear Information System (INIS)

    Herring, J.St.; Lessing, P.; O'Brien, J.E.; Stoots, C.; Hartvigsen, J.; Elangovan, S.

    2004-01-01

    An experimental research programme is being conducted by the INEEL and Ceramatec, Inc., to test the high-temperature, electrolytic production of hydrogen from steam using a solid oxide cell. The research team is designing and testing solid oxide cells for operation in the electrolysis mode, producing hydrogen rising a high-temperature heat and electrical energy. The high-temperature heat and the electrical power would be supplied simultaneously by a high-temperature nuclear reactor. Operation at high temperature reduces the electrical energy requirement for electrolysis and also increases the thermal efficiency of the power-generating cycle. The high-temperature electrolysis process will utilize heat from a specialized secondary loop carrying a steam/hydrogen mixture. It is expected that, through the combination of a high-temperature reactor and high-temperature electrolysis, the process will achieve an overall thermal conversion efficiency of 40 to 50%o while avoiding the challenging chemistry and corrosion issues associated with the thermochemical processes. Planar solid oxide cell technology is being utilised because it has the best potential for high efficiency due to minimized voltage and current losses. These losses also decrease with increasing temperature. Initial testing has determined the performance of single 'button' cells. Subsequent testing will investigate the performance of multiple-cell stacks operating in the electrolysis mode. Testing is being performed both at Ceramatec and at INEEL. The first cells to be tested were single cells based on existing materials and fabrication technology developed at Ceramatec for production of solid oxide fuel cells. These cells use a relatively thick (∼ 175 μm) electrolyte of yttria- or scandia-stabilised zirconia, with nickel-zirconia cermet anodes and strontium-doped lanthanum manganite cathodes. Additional custom cells with lanthanum gallate electrolyte have been developed and tested. Results to date have

  20. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  1. The influence of the scale effect and high temperatures on the strength and strains of high performance concrete

    Directory of Open Access Journals (Sweden)

    Korsun Vladimyr Ivanovych

    2014-03-01

    Full Text Available The most effective way to reduce the structure mass, labor input and expenses for its construction is to use modern high-performance concrete of the classes С50/60… С90/105, which possess high physical and mathematic characteristics. One of the constraints for their implementation in mass construction in Ukraine is that in design standards there are no experimental data on the physical and mathematic properties of concrete of the classes more than С50/60. Also there are no exact statements on calculating reinforced concrete structures made of high-performance concretes.The authors present the results of experimental research of the scale effect and short-term and long-term heating up to +200 ° C influence on temperature and shrinkage strain, on strength and strain characteristics under compression and tensioning of high-strength modified concrete of class C70/85. The application of high performance concretes is challenging in the process of constructing buildings aimed at operating in high technological temperatures: smoke pipes, coolers, basins, nuclear power plants' protective shells, etc. Reducing cross-sections can lead to reducing temperature drops and thermal stresses in the structures.

  2. High level nuclear wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.

    1987-01-01

    The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

  3. 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

  4. 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

  5. Metallic materials for heat exchanger components and highly stressed internal of HTR reactors for nuclear process heat generation

    International Nuclear Information System (INIS)

    1982-01-01

    The programme was aimed at the development and improvement of materials for the high-temperature heat exchanger components of a process steam HTR. The materials must have high resistance to corrosion, i.e. carburisation and internal oxidation, and high long-term toughness over a wide range of temperatures. They must also meet the requirements set in the nuclear licensing procedure, i.e. resistance to cyclic stress and irradiation, non-destructive testing, etc. Initially, it was only intended to improve and qualify commercial alloys. Later on an alloy development programme was initiated in which new, non-commercial alloys were produced and modified for use in a nuclear process heat facility. Separate abstracts were prepared for 19 pays of this volume. (orig./IHOE) [de

  6. Glass containing radioactive nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1985-01-01

    Lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level-radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800 C, since they exhibit very low melt viscosities in the 800 to 1050 C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550 C and are not adversely affected by large doses of gamma radiation in H 2 O at 135 C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms. (author)

  7. Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

    Energy Technology Data Exchange (ETDEWEB)

    Kim Davies; Shelly X Li

    2007-09-01

    Pyrochemical processing plays an important role in development of proliferation- resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ~2 grams of LiCl/KCl salt electrolyte with a low concentration (~1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver wire

  8. Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

    International Nuclear Information System (INIS)

    Kim Davies; Shelly X Li

    2007-01-01

    Pyrochemical processing plays an important role in development of proliferation-resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ∼2 grams of LiCl/KCl salt electrolyte with a low concentration (∼1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver

  9. Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this project is to develop a fundamental understanding of the mechanisms that limit materials durability for very high-temperature applications. Current design limitations are based on material strength and corrosion resistance. This project will characterize the interactions of high-temperature creep, fatigue, and environmental attack in structural metallic alloys of interest for the very high-temperature gas-cooled reactor (VHTR) or Next Generation Nuclear Plant (NGNP) and for the associated thermo-chemical processing systems for hydrogen generation. Each of these degradation processes presents a major materials design challenge on its own, but in combination, they can act synergistically to rapidly degrade materials and limit component lives. This research and development effort will provide experimental results to characterize creep-fatigue-environment interactions and develop predictive models to define operation limits for high-temperature structural material applications. Researchers will study individually and in combination creep-fatigue-environmental attack processes in Alloys 617, 230, and 800H, as well as in an advanced Ni-Cr oxide dispersion strengthened steel (ODS) system. For comparison, the study will also examine basic degradation processes in nichrome (Ni-20Cr), which is a basis for most high-temperature structural materials, as well as many of the superalloys. These materials are selected to represent primary candidate alloys, one advanced developmental alloy that may have superior high-temperature durability, and one model system on which basic performance and modeling efforts can be based. The research program is presented in four parts, which all complement each other. The first three are primarily experimental in nature, and the last will tie the work together in a coordinated modeling effort. The sections are (1) dynamic creep-fatigue-environment process, (2) subcritical crack processes, (3) dynamic corrosion crack

  10. Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James; Gewirth, Andrew; Sehitoglu, Huseyin; Sofronis, Petros; Robertson, Ian

    2014-01-16

    The objective of this project is to develop a fundamental understanding of the mechanisms that limit materials durability for very high-temperature applications. Current design limitations are based on material strength and corrosion resistance. This project will characterize the interactions of high-temperature creep, fatigue, and environmental attack in structural metallic alloys of interest for the very high-temperature gas-cooled reactor (VHTR) or Next–Generation Nuclear Plant (NGNP) and for the associated thermo-chemical processing systems for hydrogen generation. Each of these degradation processes presents a major materials design challenge on its own, but in combination, they can act synergistically to rapidly degrade materials and limit component lives. This research and development effort will provide experimental results to characterize creep-fatigue-environment interactions and develop predictive models to define operation limits for high-temperature structural material applications. Researchers will study individually and in combination creep-fatigue-environmental attack processes in Alloys 617, 230, and 800H, as well as in an advanced Ni-Cr oxide dispersion strengthened steel (ODS) system. For comparison, the study will also examine basic degradation processes in nichrome (Ni-20Cr), which is a basis for most high-temperature structural materials, as well as many of the superalloys. These materials are selected to represent primary candidate alloys, one advanced developmental alloy that may have superior high-temperature durability, and one model system on which basic performance and modeling efforts can be based. The research program is presented in four parts, which all complement each other. The first three are primarily experimental in nature, and the last will tie the work together in a coordinated modeling effort. The sections are (1) dynamic creep-fatigue-environment process, (2) subcritical crack processes, (3) dynamic corrosion – crack

  11. High temperature structural silicides

    International Nuclear Information System (INIS)

    Petrovic, J.J.

    1997-01-01

    Structural silicides have important high temperature applications in oxidizing and aggressive environments. Most prominent are MoSi 2 -based materials, which are borderline ceramic-intermetallic compounds. MoSi 2 single crystals exhibit macroscopic compressive ductility at temperatures below room temperature in some orientations. Polycrystalline MoSi 2 possesses elevated temperature creep behavior which is highly sensitive to grain size. MoSi 2 -Si 3 N 4 composites show an important combination of oxidation resistance, creep resistance, and low temperature fracture toughness. Current potential applications of MoSi 2 -based materials include furnace heating elements, molten metal lances, industrial gas burners, aerospace turbine engine components, diesel engine glow plugs, and materials for glass processing

  12. High temperature resistive phase transition in A15 high temperature superconductors

    International Nuclear Information System (INIS)

    Chu, C.W.; Huang, C.Y.; Schmidt, P.H.; Sugawara, K.

    1976-01-01

    Resistive measurements were made on A15 high temperature superconductors. Anomalies indicative of a phase transition were observed at 433 0 K in a single crystal Nb 3 Sn and at 485 0 K in an unbacked Nb 3 Ge sputtered thin film. Results are compared with the high temperature transmission electron diffraction studies of Nb 3 Ge films by Schmidt et al. A possible instability in the electron energy spectrum is discussed

  13. Development of high temperature property database for Alloy 800H

    International Nuclear Information System (INIS)

    Yokoyama, Norio; Watanabe, Katsutoshi; Tsuji, Hirokazu; Nakajima, Hajime.

    1993-07-01

    JAERI Material Performance Database (JMPD) has been developed since 1989 in JAERI with a view to utilizing the various kinds of characteristic data of nuclear materials efficiently. Using relational database management system, PLANNER on the mainframe, the JMPD provides the retrieval supporting system, graphic and statistical analyses system. The data obtained with 7868 sets on characteristic data of metallic materials including fatigue crack growth data, etc. have been stored in the JMPD at the end of March in 1993. A ferritic superalloy, Alloy 800H is used for the structural material of the control rods of the High Temperature Engineering Test Reactor (HTTR). Thermal stress generates which might cause a severe creep damage at a reactor scram. It therefore needs to be designed with consideration on the fracture modes induced by creep deformation after neutron irradiation. The creep data (approximately 240 sets) and tensile data (approximately 100 sets) of Alloy 800H including the effects of test environment, aging treatment and neutron irradiation have been stored in the JMPD. Furthermore, using a personal computer, high temperature property database for Alloy 800H has been developed. The present report outlines the development of high temperature property database for Alloy 800H. (author)

  14. Full-Length High-Temperature Severe Fuel Damage Test No. 5: Final safety analysis

    International Nuclear Information System (INIS)

    Lanning, D.D.; Lombardo, N.J.; Panisko, F.E.

    1993-09-01

    This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage

  15. Hydrogen production methods efficiency coupled to an advanced high temperature accelerator driven system

    International Nuclear Information System (INIS)

    Rodríguez, Daniel González; Lira, Carlos Alberto Brayner de Oliveira

    2017-01-01

    The hydrogen economy is one of the most promising concepts for the energy future. In this scenario, oil is replaced by hydrogen as an energy carrier. This hydrogen, rather than oil, must be produced in volumes not provided by the currently employed methods. In this work two high temperature hydrogen production methods coupled to an advanced nuclear system are presented. A new design of a pebbled-bed accelerator nuclear driven system called TADSEA is chosen because of the advantages it has in matters of transmutation and safety. For the conceptual design of the high temperature electrolysis process a detailed computational fluid dynamics model was developed to analyze the solid oxide electrolytic cell that has a huge influence on the process efficiency. A detailed flowsheet of the high temperature electrolysis process coupled to TADSEA through a Brayton gas cycle was developed using chemical process simulation software: Aspen HYSYS®. The model with optimized operating conditions produces 0.1627 kg/s of hydrogen, resulting in an overall process efficiency of 34.51%, a value in the range of results reported by other authors. A conceptual design of the iodine-sulfur thermochemical water splitting cycle was also developed. The overall efficiency of the process was calculated performing an energy balance resulting in 22.56%. The values of efficiency, hydrogen production rate and energy consumption of the proposed models are in the values considered acceptable in the hydrogen economy concept, being also compatible with the TADSEA design parameters. (author)

  16. Hydrogen production methods efficiency coupled to an advanced high temperature accelerator driven system

    Energy Technology Data Exchange (ETDEWEB)

    Rodríguez, Daniel González; Lira, Carlos Alberto Brayner de Oliveira [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Fernández, Carlos García, E-mail: danielgonro@gmail.com, E-mail: mmhamada@ipen.br [Instituto Superior de Tecnologías y Ciencias aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The hydrogen economy is one of the most promising concepts for the energy future. In this scenario, oil is replaced by hydrogen as an energy carrier. This hydrogen, rather than oil, must be produced in volumes not provided by the currently employed methods. In this work two high temperature hydrogen production methods coupled to an advanced nuclear system are presented. A new design of a pebbled-bed accelerator nuclear driven system called TADSEA is chosen because of the advantages it has in matters of transmutation and safety. For the conceptual design of the high temperature electrolysis process a detailed computational fluid dynamics model was developed to analyze the solid oxide electrolytic cell that has a huge influence on the process efficiency. A detailed flowsheet of the high temperature electrolysis process coupled to TADSEA through a Brayton gas cycle was developed using chemical process simulation software: Aspen HYSYS®. The model with optimized operating conditions produces 0.1627 kg/s of hydrogen, resulting in an overall process efficiency of 34.51%, a value in the range of results reported by other authors. A conceptual design of the iodine-sulfur thermochemical water splitting cycle was also developed. The overall efficiency of the process was calculated performing an energy balance resulting in 22.56%. The values of efficiency, hydrogen production rate and energy consumption of the proposed models are in the values considered acceptable in the hydrogen economy concept, being also compatible with the TADSEA design parameters. (author)

  17. Hydrogen energy based on nuclear energy

    International Nuclear Information System (INIS)

    2002-06-01

    A concept to produce hydrogen of an energy carrier using nuclear energy was proposed since 1970s, and a number of process based on thermochemical method has been investigated after petroleum shock. As this method is used high temperature based on nuclear reactors, these researches are mainly carried out as a part of application of high temperature reactors, which has been carried out at an aim of the high temperature reactor application in the Japan Atomic Energy Research Institute. On October, 2000, the 'First International Conference for Information Exchange on Hydrogen Production based on Nuclear Energy' was held by auspice of OECD/NEA, where hydrogen energy at energy view in the 21st Century, technology on hydrogen production using nuclear energy, and so on, were published. This commentary was summarized surveys and researches on hydrogen production using nuclear energy carried out by the Nuclear Hydrogen Research Group established on January, 2001 for one year. They contains, views on energy and hydrogen/nuclear energy, hydrogen production using nuclear energy and already finished researches, methods of hydrogen production using nuclear energy and their present conditions, concepts on production plants of nuclear hydrogen, resources on nuclear hydrogen production and effect on global environment, requests from market and acceptability of society, and its future process. (G.K.)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. A protection system of low temperature thermo-supply nuclear reactor

    International Nuclear Information System (INIS)

    Jiang Binsen

    1988-09-01

    A Protection system of low temperature thermo-supply nuclear reactor is introduced. It is the first protection system, which is designed and manufactred on the basis of Chinese National Standard GB 4083-83 'General Safety Principle of Nuclear Reactor Protection System', to be considered under the circumstances of industry level in China. Advantages of the protection system are as follows: 1)The single failure criteria can fully be fulfilled by the protection system. 2) On-line testing system can be used for detecting all of failure components and quick identifying the failure points in the system. 3) It is convenience for maintenacnce of the system. To complete this project is very important and helpful in promoting the development of the protection system and safety operation of nuclear reactor in China

  4. Pressure Resistance Welding of High Temperature Metallic Materials

    International Nuclear Information System (INIS)

    Jerred, N.; Zirker, L.; Charit, I.; Cole, J.; Frary, M.; Butt, D.; Meyer, M.; Murty, K.L.

    2010-01-01

    Pressure Resistance Welding (PRW) is a solid state joining process used for various high temperature metallic materials (Oxide dispersion strengthened alloys of MA957, MA754; martensitic alloy HT-9, tungsten etc.) for advanced nuclear reactor applications. A new PRW machine has been installed at the Center for Advanced Energy Studies (CAES) in Idaho Falls for conducting joining research for nuclear applications. The key emphasis has been on understanding processing-microstructure-property relationships. Initial studies have shown that sound joints can be made between dissimilar materials such as MA957 alloy cladding tubes and HT-9 end plugs, and MA754 and HT-9 coupons. Limited burst testing of MA957/HT-9 joints carried out at various pressures up to 400 C has shown encouraging results in that the joint regions do not develop any cracking. Similar joint strength observations have also been made by performing simple bend tests. Detailed microstructural studies using SEM/EBSD tools and fatigue crack growth studies of MA754/HT-9 joints are ongoing.

  5. High-entropy alloys as high-temperature thermoelectric materials

    Energy Technology Data Exchange (ETDEWEB)

    Shafeie, Samrand [Surface and Microstructure Engineering Group, Materials and Manufacturing Technology, Chalmers University of Technology, SE-41296 Gothenburg (Sweden); Department of Chemistry and Chemical Engineering, Chalmers University of Technology, SE-41296 Gothenburg (Sweden); Guo, Sheng, E-mail: sheng.guo@chalmers.se [Surface and Microstructure Engineering Group, Materials and Manufacturing Technology, Chalmers University of Technology, SE-41296 Gothenburg (Sweden); Hu, Qiang [Institute of Applied Physics, Jiangxi Academy of Sciences, Nanchang 330029 (China); Fahlquist, Henrik [Bruker AXS Nordic AB, 17067 Solna (Sweden); Erhart, Paul [Department of Applied Physics, Chalmers University of Technology, SE-41296 Gothenburg (Sweden); Palmqvist, Anders, E-mail: anders.palmqvist@chalmers.se [Department of Chemistry and Chemical Engineering, Chalmers University of Technology, SE-41296 Gothenburg (Sweden)

    2015-11-14

    Thermoelectric (TE) generators that efficiently recycle a large portion of waste heat will be an important complementary energy technology in the future. While many efficient TE materials exist in the lower temperature region, few are efficient at high temperatures. Here, we present the high temperature properties of high-entropy alloys (HEAs), as a potential new class of high temperature TE materials. We show that their TE properties can be controlled significantly by changing the valence electron concentration (VEC) of the system with appropriate substitutional elements. Both the electrical and thermal transport properties in this system were found to decrease with a lower VEC number. Overall, the large microstructural complexity and lower average VEC in these types of alloys can potentially be used to lower both the total and the lattice thermal conductivity. These findings highlight the possibility to exploit HEAs as a new class of future high temperature TE materials.

  6. Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

    International Nuclear Information System (INIS)

    Steven R. Sherman

    2007-01-01

    The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant

  7. High Committee for transparency and information on nuclear safety. Meeting of the High Committee on the 15 September 2011. Plenary meeting

    International Nuclear Information System (INIS)

    2011-01-01

    After the opinion of the High Committee on public inquiries, several Power Point contributions are proposed. They respectively propose a presentation of the itinerant exhibition 'Radioactivity from Homer to Oppenheimer', an assessment of nuclear material stocks and flows for 2009 and 2010 (depleted uranium, recycled uranium, plutonium, uranium and plutonium exports, materials generated by the supply of the EDF fleet), a presentation of the Generation IV International Forum (with a presentation of the six concepts for fourth generation nuclear reactor: sodium fast neutrons, lead fast neutrons, gas fast neutrons, very high temperature and gas, super-critical water, molten salts), a discussion on the interest of fourth generation reactors and their associated fuel cycle (the example of ASTRID), several discussions (by EDF, by the ASN, by the IRSN) on an operation accident in the CENTRACO base nuclear installation in September 2011), and a description of the participation of the Gravelines Local Information Commission to the nuclear crisis exercise in January 2011

  8. Uranium-thorium fuel cycle in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

    2011-01-01

    Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

  9. The on-line low temperature nuclear orientation facility NICOLE

    International Nuclear Information System (INIS)

    Ohtsubo, T; Roccia, S; Gaulard, C; Stone, N J; Stone, J R; Köster, U; Nikolov, J; Veskovic, M; Simpson, G S

    2017-01-01

    We review major experiments and results obtained by the on-line low temperature nuclear orientation method at the NICOLE facility at ISOLDE, CERN since the year 2000 and highlight their general physical impact. This versatile facility, providing a large degree of controlled nuclear polarization, was used for a long-standing study of magnetic moments at shell closures in the region Z  = 28, N  = 28–50 but also for dedicated studies in the deformed region around A  ∼ 180. Another physics program was conducted to test symmetry in the weak sector and constrain weak coupling beyond V–A . Those two programs were supported by careful measurements of the involved solid state physics parameters to attain the full sensitivity of the technique and provide interesting interdisciplinary results. Future plans for this facility include the challenging idea of measuring the beta–gamma–neutron angular distributions from polarized beta delayed neutron emitters, further test of fundamental symmetries and obtaining nuclear structure data used in medical applications. The facility will also continue to contribute to both the nuclear structure and fundamental symmetry test programs. (paper)

  10. On the nuclear fragmentation mechanisms in nuclear collisions at intermediate and high energies

    International Nuclear Information System (INIS)

    Jipa, Al; Besliu, C.; Felea, D.

    2004-01-01

    The nuclear fragmentation mechanisms can be discussed by taking into account different scales related to the fragment sizes. Considering two fragmentation mechanisms of the nuclei at the same incident energy an analysis of the experimental results obtained was done. Goldhaber formula was improved by analyzing the discrepancies between data and theories concerning the projectile fragmentation. We implied that the projectile fragmentation process would be governed by the distribution of nucleon momenta in the projectile after the collision occurred. We used in our analysis protons from the 4 He + 7 Li at 4.5 GeV/c per nucleon incident momentum, as well as from 40 Ar + 12 C at 213 AMeV bombarding energy. We proved that in order to proceed in analyzing the projectile fragmentation process at intermediate and high energies one has to consider the dependence σ 0 on the apparent temperature of projectile nucleus after the collision took place. The generalized Bertsch correction for light projectile nuclei and fragments was used and the number of spatial correlations between identical nucleons having anticorrelated momenta was found. Thus we found apparent temperature values close to the separation energies of the considered fragments per number of fragments. The temperatures associated to kinetic energy spectra of the projectile fragments were calculated following two methods. The results from Bauer's method were compared with those obtained by fitting the kinetic energy distributions of the projectile fragments in the rest frame of the projectile with a Maxwellian curve. We also accomplished the comparison of the experimental results with similar events simulated with RQMD 2.4. All the results obtained suggested two nuclear fragmentation mechanisms: a sudden fragmentation by explosive mechanisms, like shock waves and a slow fragmentation by the 'fission' of the spectator regions, mainly because of the interactions with the particles or fragments emitted from the

  11. 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

  12. 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

  13. Development status on hydrogen production technology using high-temperature gas-cooled reactor at JAEA, Japan

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

    2006-01-01

    The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR (up to 900 deg. C) offers one of the most promising technological solutions to curb the rising level of CO 2 emission and resulting risk of climate change. The interests in HTGR as an advanced nuclear power source for the next generation reactor, therefore, continue to rise. This is represented by the Japanese HTTR (High-Temperature Engineering Test Reactor) Project and the Chinese HTR-10 Project, followed by the international Generation IV development program, US nuclear hydrogen initiative program, EU innovative HTR technology development program, etc. To enhance nuclear energy application to heat process industries, the Japan Atomic Energy Agency (JAEA) has continued extensive efforts for development of hydrogen production system using the nuclear heat from HTGR in the framework of the HTTR Project. The HTTR Project has the objectives of establishing both HTGR technology and heat utilization technology. Using the HTTR constructed at the Oarai Research and Development Center of JAEA, reactor performance and safety demonstration tests have been conducted as planned. The reactor outlet temperature of 950 deg. C was successfully achieved in April 2004. For hydrogen production as heat utilization technology, R and D on thermo-chemical water splitting by the 'Iodine-Sulfur process' (IS process) has been conducted step by step. Proof of the basic IS process was made in 1997 on a lab-scale of hydrogen production of 1 L/h. In 2004, one-week continuous operation of the IS process was successfully demonstrated using a bench-scale apparatus with hydrogen production rate of 31 L/h. Further test using a pilot scale facility with greater hydrogen production rate of 10 - 30 m 3 /h is planned as

  14. High resolution spectroscopy in solids by nuclear magnetic resonance

    International Nuclear Information System (INIS)

    Bonagamba, T.J.

    1991-07-01

    The nuclear magnetic resonance (NMR) techniques for High Resolution Spectroscopy in Solids are described. Also the construction project of a partially home made spectrometer and its applications in the characterization of solid samples are shown in detail. The high resolution spectrometer used is implemented with the double resonance multiple pulses sequences and magic angle spinning (MAS) and can be used with solid and liquid samples. The maximum spinning frequency for the MAS experiment is in excess of 5 Khz, the double resonance sequences can be performed with any type of nucleus, in the variable temperature operating range with nitrogen gas: -120 0 C to +160 0 C, and is fully controlled by a Macintosh IIci microcomputer. (author)

  15. Papers about coated particles, graphitic and metallic materials for progressive high-temperature reactors at the Reactor Meeting 1978

    International Nuclear Information System (INIS)

    Rottmann, J.

    1978-09-01

    In the contributions, questions on the development, the radiation and the high-temperature behaviour and the characterization of fuel element particles are treated. Furthermore the resistance and radiation behaviour of graphitic materials are discussed. Finally, questions on the choice of high-temperature alloys for nuclear process heat facilities are discussed and the testing-equipment of the Nuclear Research Centre as well as first results of the long-time experiments are presented. The work was performed within the frame of the projects 'HTR-Fuel Element Cycle' and 'Prototype Nuclear Process Heat', which are sponsored by the Federal Ministry of Research and Technology of the Federal Republic of Germany and of the state of North-Rhine-Westfalia. Partner firms, who participate in the two projects are Gelsenberg AG, Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelement GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, SIGRI Elektrographit GmbH/Ringsdorff-Werke GmbH, Bergbauforschung GmbH und Rheinische Braunkohlenwerke AG. (orig./UA) [de

  16. High temperature vapors science and technology

    CERN Document Server

    Hastie, John

    2012-01-01

    High Temperature Vapors: Science and Technology focuses on the relationship of the basic science of high-temperature vapors to some areas of discernible practical importance in modern science and technology. The major high-temperature problem areas selected for discussion include chemical vapor transport and deposition; the vapor phase aspects of corrosion, combustion, and energy systems; and extraterrestrial high-temperature species. This book is comprised of seven chapters and begins with an introduction to the nature of the high-temperature vapor state, the scope and literature of high-temp

  17. 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

  18. Evaluation of high temperature pressure sensors

    International Nuclear Information System (INIS)

    Choi, In-Mook; Woo, Sam-Yong; Kim, Yong-Kyu

    2011-01-01

    It is becoming more important to measure the pressure in high temperature environments in many industrial fields. However, there is no appropriate evaluation system and compensation method for high temperature pressure sensors since most pressure standards have been established at room temperature. In order to evaluate the high temperature pressure sensors used in harsh environments, such as high temperatures above 250 deg. C, a specialized system has been constructed and evaluated in this study. The pressure standard established at room temperature is connected to a high temperature pressure sensor through a chiller. The sensor can be evaluated in conditions of changing standard pressures at constant temperatures and of changing temperatures at constant pressures. According to the evaluation conditions, two compensation methods are proposed to eliminate deviation due to sensitivity changes and nonlinear behaviors except thermal hysteresis.

  19. Where is high technology taking nuclear medicine

    International Nuclear Information System (INIS)

    Veall, N.

    1985-01-01

    The question is posed as to whether high technology in nuclear medicine might lead to the nuclear medicine practitioner possibly finishing up working for the machine rather than the improvement of health care in its widest sense. A brief examination of some pros and cons of high technology nuclear medicine is given. (U.K.)

  20. High-Frequency Dynamic Nuclear Polarization in the Nuclear Rotating Frame

    DEFF Research Database (Denmark)

    Farrar, C. T.; Hall, D. A.; Gerfen, G. J.

    2000-01-01

    A proton dynamic nuclear polarization (DNP) NMR signal enhancement (ϵ) close to thermal equilibrium, ϵ = 0.89, has been obtained at high field (B0 = 5 T, νepr = 139.5 GHz) using 15 mM trityl radical in a 40:60 water/glycerol frozen solution at 11 K. The electron-nuclear polarization transfer...... is performed in the nuclear rotating frame with microwave irradiation during a nuclear spin-lock pulse. The growth of the signal enhancement is governed by the rotating frame nuclear spin–lattice relaxation time (T1ρ), which is four orders of magnitude shorter than the nuclear spin–lattice relaxation time (T1n......). Due to the rapid polarization transfer in the nuclear rotating frame the experiment can be recycled at a rate of 1/T1ρ and is not limited by the much slower lab frame nuclear spin–lattice relaxation rate (1/T1n). The increased repetition rate allowed in the nuclear rotating frame provides an effective...

  1. 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.

  2. 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

  3. Issues of high-burnup fuel for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Belac, J.; Milisdoerfer, L.

    2004-12-01

    A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

  4. New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M.; Vlahovic, L.; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe, (Germany); Sadli, M.; Failleau, G. [Laboratoire Commun de Metrologie, LNE-Cnam, Saint-Denis, (France); Fuetterer, M.; Lapetite, J.M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, (Netherlands); Fourrez, S. [Thermocoax, 8 rue du pre neuf, F-61100 St Georges des Groseillers, (France)

    2015-07-01

    Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand high temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)

  5. The determination of nuclear matter temperature and density

    International Nuclear Information System (INIS)

    Wolf, K.L.

    1981-01-01

    The purpose of this paper is to review some of the things we have learned about nuclear matter under extreme conditions during the past few years in relativistic heavy ion studies. High energy heavy-ion collisions provide a unique mechanism for exploring the dependence of the nuclear potential energy epsilon(rho,T) on the degree of compression and excitation, and may even show the existence of new phases of matter. Thus the determination of the nuclear equation of state remains the ultimate goal of many researchers in this field. (orig.)

  6. 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.

  7. A High Temperature Liquid Plasma Model of the Sun

    Directory of Open Access Journals (Sweden)

    Robitaille P.-M.

    2007-01-01

    Full Text Available In this work, a liquid model of the Sun is presented wherein the entire solar mass is viewed as a high density/high energy plasma. This model challenges our current understanding of the densities associated with the internal layers of the Sun, advocating a relatively constant density, almost independent of radial position. The incompressible nature of liquids is advanced to prevent solar collapse from gravitational forces. The liquid plasma model of the Sun is a non-equilibrium approach, where nuclear reactions occur throughout the solar mass. The primary means of addressing internal heat transfer are convection and conduction. As a result of the convective processes on the solar surface, the liquid model brings into question the established temperature of the solar photosphere by highlighting a violation of Kirchhoff’s law of thermal emission. Along these lines, the model also emphasizes that radiative emission is a surface phenomenon. Evidence that the Sun is a high density/high energy plasma is based on our knowledge of Planckian thermal emission and condensed matter, including the existence of pressure ionization and liquid metallic hydrogen at high temperatures and pressures. Prior to introducing the liquid plasma model, the historic and scientific justifications for the gaseous model of the Sun are reviewed and the gaseous equations of state are also discussed.

  8. Theoretical interpretation of high-energy nuclear collisions

    International Nuclear Information System (INIS)

    Fai, G.

    1992-06-01

    Nuclear collisions are interpreted theoretically. The nuclear equation of state is studied in a wide energy range. Subnucleonic degrees of freedom are invoked at high energy densities and at short length-scales. Questions of dynamical collision simulations are investigated. Direct support is provided for experiment in the form of collaborative projects. The major objective of this nuclear theory program is a better understanding of the properties of strongly interacting matter on the nuclear energy scale, as manifested in high-energy heavy-ion collisions

  9. Nuclear grade cable thermal life model by time temperature superposition algorithm based on Matlab GUI

    International Nuclear Information System (INIS)

    Lu Yanyun; Gu Shenjie; Lou Tianyang

    2014-01-01

    Background: As nuclear grade cable must endure harsh environment within design life, it is critical to predict cable thermal life accurately owing to thermal aging, which is one of dominant factors of aging mechanism. Purpose: Using time temperature superposition (TTS) method, the aim is to construct nuclear grade cable thermal life model, predict cable residual life and develop life model interactive interface under Matlab GUI. Methods: According to TTS, nuclear grade cable thermal life model can be constructed by shifting data groups at various temperatures to preset reference temperature with translation factor which is determined by non linear programming optimization. Interactive interface of cable thermal life model developed under Matlab GUI consists of superposition mode and standard mode which include features such as optimization of translation factor, calculation of activation energy, construction of thermal aging curve and analysis of aging mechanism., Results: With calculation result comparison between superposition and standard method, the result with TTS has better accuracy than that with standard method. Furthermore, confidence level of nuclear grade cable thermal life with TTS is higher than that with standard method. Conclusion: The results show that TTS methodology is applicable to thermal life prediction of nuclear grade cable. Interactive Interface under Matlab GUI achieves anticipated functionalities. (authors)

  10. Low-Temperature Dynamic Nuclear Polarization at 9.4 Tesla With a 30 Milliwatt Microwave Source

    Science.gov (United States)

    Thurber, Kent R.; Yau, Wai-Ming; Tycko, Robert

    2010-01-01

    Dynamic nuclear polarization (DNP) can provide large signal enhancements in nuclear magnetic resonance (NMR) by transfer of polarization from electron spins to nuclear spins. We discuss several aspects of DNP experiments at 9.4 Tesla (400 MHz resonant frequency for 1H, 264 GHz for electron spins in organic radicals) in the 7–80 K temperature range, using a 30 mW, frequency-tunable microwave source and a quasi-optical microwave bridge for polarization control and low-loss microwave transmission. In experiments on frozen glycerol/water doped with nitroxide radicals, DNP signal enhancements up to a factor of 80 are observed (relative to 1H NMR signals with thermal equilibrium spin polarization). The largest sensitivity enhancements are observed with a new triradical dopant, DOTOPA-TEMPO. Field modulation with a 10 G root-mean-squared amplitude during DNP increases the nuclear spin polarizations by up to 135%. Dependencies of 1H NMR signal amplitudes, nuclear spin relaxation times, and DNP build-up times on the dopant and its concentration, temperature, microwave power, and modulation frequency are reported and discussed. The benefits of low-temperature DNP can be dramatic: the 1H spin polarization is increased approximately 1000-fold at 7 K with DNP, relative to thermal polarization at 80 K. PMID:20392658

  11. 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)

  12. 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)

  13. 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

  14. Study of elastic and thermodynamic properties of uranium dioxide under high temperature and pressure with density functional theory

    International Nuclear Information System (INIS)

    Zhou Mu; Wang Feng; Zheng Zhou; Liu Xiankun; Jiang Tao

    2013-01-01

    The elastic and thermodynamic properties of UO 2 under extreme physical condition are studied by using the density functional theory and quasi-harmonic Debye model. Results show that UO 2 is still stable ionic crystal under high temperatures, and pressures. Tetragonal shear constant is steady under high pressures and temperatures, while elastic constant C 44 is stable under high temperatures, but rises with pressure sharply. Bulk modulus, shear modulus and Young's modulus increase with pressure rapidly, but temperature would not cause evident debasement of the moduli, all of which indicate that UO 2 has excellent mechanical properties. Heat capacity of different pressures increases with temperature and is close to the Dulong-Petit limit near 1000 K. Debye temperature decreases with temperature, and increases with pressure. Under low pressure, thermal expansion coefficient raises with temperature rapidly, and then gets slow at higher pressure and temperature. Besides, the thermal expansion coefficient of UO 2 is much lower than that of other nuclear materials. (authors)

  15. High-temperature thermodynamic data for species in aqueous solution. Final report

    International Nuclear Information System (INIS)

    Cobble, J.W.; Murray, R.C. Jr.; Turner, P.J.; Chen, K.

    1982-05-01

    This final report summarizes the results of experimental and theoretical research on the high temperature thermodynamic properties of aqueous species important to nuclear reactor water chemistry. Methods of predicting thermodynamic functions are included for electrolytes up to 300 0 C where experimental data are lacking. Data in the literature are evaluated and tables of important equilibrium constants for 78 reactions encountered in corrosion and precipitation in nuclear reactors are listed up to 300 0 C. Finally, tables of free energy functions from 0 to 300 0 C are given for 56 individual species. These data represented form a major compilation resulting from the most advanced experimental and theoretical methods. Illustrations of the use of the tables are given for problems involving pH control, precipitation, and corrosion. 11 figures, 100 tables

  16. Use of aluminum nitride to obtain temperature measurements in a high temperature and high radiation environment

    Science.gov (United States)

    Wernsman, Bernard R.; Blasi, Raymond J.; Tittman, Bernhard R.; Parks, David A.

    2016-04-26

    An aluminum nitride piezoelectric ultrasonic transducer successfully operates at temperatures of up to 1000.degree. C. and fast (>1 MeV) neutron fluencies of more than 10.sup.18 n/cm.sup.2. The transducer comprises a transparent, nitrogen rich aluminum nitride (AlN) crystal wafer that is coupled to an aluminum cylinder for pulse-echo measurements. The transducer has the capability to measure in situ gamma heating within the core of a nuclear reactor.

  17. Transverse momentum in high-energy nuclear collisions: Collective expansion

    International Nuclear Information System (INIS)

    Wang, X.; Hwa, R.C.

    1987-01-01

    Hadron production in the central region in high-energy nuclear collisions is investigated. The hydrodynamical expansion of a locally thermalized system is studied for both the cases with and without phase transition. The case with phase transition is considered by using a sound-velocity function c/sub s/(T) parametrized to fit the energy density determined in a lattice gauge calculation. The effect of a transverse rarefaction wave is included in the calculation of the temperature profile of the expanding fluid. The transverse-momentum distribution of hadrons is calculated by collecting all the hadrons produced when the hadron gas is cooled down to a freeze-out temperature at different times in the expansion. Fluctuation in initial temperature and radius is allowed due to variation in impact parameter. On the basis of a study of the thermalization process in the parton model we impose a constraint on the initial temperature and the thermalization time, the simultaneous variation of both of which gives rise to a relationship between the average transverse momentum and rapidity density. We have found that there is no so-called ''plateau'' region in that relationship. The implication on the diagnostics of a quark-gluon plasma is discussed

  18. Water cooled type nuclear power plant

    International Nuclear Information System (INIS)

    Arai, Shigeki.

    1981-01-01

    Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

  19. 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)

  20. Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Matt Ebner; Manohar Sohal; Phil Sharpe; Thermal Hydraulics Group

    2010-03-01

    Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

  1. Theoretical interpretation of high-energy nuclear collisions

    International Nuclear Information System (INIS)

    Fai, G.

    1991-07-01

    Nuclear collision data are interpreted theoretically. The nuclear equation of state is investigated with particular emphasis on momentum-dependent mean field effects. Subnucleonic degrees of freedom are invoked at high energies and densities, and a short length-scales. A nontopological soliton model for baryons is studied in which effective meson fields are generated from extended quark-antiquark pairs. The major objective of this nuclear theory project is a better understanding of the properties of strongly interacting matter on the nuclear energy scale, as manifested in high-energy heavy-ion collisions

  2. Quantum electrodynamics at high temperature. 2

    International Nuclear Information System (INIS)

    Alvarez-Estrada, R.F.

    1988-01-01

    The photon sector of QED in d = 3 spatial dimensions is analyzed at high temperature thereby generalizing nontrivially a previous study for d = 1. The imaginary time formalism and an improved renormalized perturbation theory which incorporates second order Debye screening are used. General results are presented for the leading high temperature contributions to all renormalized connected photon Green's functions for fixed external momenta (much smaller than the temperature) to all orders in the improved perturbation theory. Those leading contributions are ultraviolet finite, infrared convergent and gauge invariant, and display an interesting form of dimensional reduction at high temperature. A new path integral representations is given for the high temperature partition function with an external photon source, which is shown to generate all leading high temperature Green's functions mentioned above, and, so, it displays neatly the kind of dimensional reduction which makes QED to become simpler at high temperature. This limiting partition function corresponds to an imaginary time dependent electron positron field interacting with an electromagnetic field at zero imaginary time, and it depends on the renormalized electron mass and electric charge, the second order contribution to the usual renormalization constant Z 3 and a new mass term, which is associated to the photon field with vanishing Lorentz index. The new mass term corresponds to a finite number of diagrams in the high temperature improved perturbation theory and carriers ultraviolet divergences which are compensated for by other contributions (so that the leading high temperature Green's functions referred to above are ultraviolet finite). The dominant high temperature contributions to the renormalized thermodynamic potential to all perturbative orders: i) are given in terms of the above leading high-temperature contributions to the photon Green's functions (except for a few diagrams of low order in the

  3. Thermal control of high energy nuclear waste, space option. [mathematical models

    Science.gov (United States)

    Peoples, J. A.

    1979-01-01

    Problems related to the temperature and packaging of nuclear waste material for disposal in space are explored. An approach is suggested for solving both problems with emphasis on high energy density waste material. A passive cooling concept is presented which utilized conduction rods that penetrate the inner core. Data are presented to illustrate the effectiveness of the rods and the limit of their capability. A computerized thermal model is discussed and developed for the cooling concept.

  4. 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

  5. Intermediate/high energy nuclear physics

    International Nuclear Information System (INIS)

    Vary, J.P.

    1992-01-01

    Progress during the last year is reviewed under the following topics: relativistic hadron--nucleus and nucleus--nucleus collisions (heavy meson production, photon production and fragmentation functions--direct photon production with the QCM and photon fragmentation functions, Cronin efffect and multiple scattering, effective nuclear parton distributions); solving quantum field theories in nonperturbative regime; light-front dynamics and high-spin states (soft form factor of the pion and nucleon for transverse and longitudinal momentum transfers, light front spinors for high-spin objects); high-energy spin physics; relativistic wave equations, quarkonia, and e + e - resonances; associated production of Higgs boson at collider energies, and microscopic nuclear many-body theory and reactions. 135 refs

  6. Passive Resistor Temperature Compensation for a High-Temperature Piezoresistive Pressure Sensor.

    Science.gov (United States)

    Yao, Zong; Liang, Ting; Jia, Pinggang; Hong, Yingping; Qi, Lei; Lei, Cheng; Zhang, Bin; Li, Wangwang; Zhang, Diya; Xiong, Jijun

    2016-07-22

    The main limitation of high-temperature piezoresistive pressure sensors is the variation of output voltage with operating temperature, which seriously reduces their measurement accuracy. This paper presents a passive resistor temperature compensation technique whose parameters are calculated using differential equations. Unlike traditional experiential arithmetic, the differential equations are independent of the parameter deviation among the piezoresistors of the microelectromechanical pressure sensor and the residual stress caused by the fabrication process or a mismatch in the thermal expansion coefficients. The differential equations are solved using calibration data from uncompensated high-temperature piezoresistive pressure sensors. Tests conducted on the calibrated equipment at various temperatures and pressures show that the passive resistor temperature compensation produces a remarkable effect. Additionally, a high-temperature signal-conditioning circuit is used to improve the output sensitivity of the sensor, which can be reduced by the temperature compensation. Compared to traditional experiential arithmetic, the proposed passive resistor temperature compensation technique exhibits less temperature drift and is expected to be highly applicable for pressure measurements in harsh environments with large temperature variations.

  7. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  8. Microstructural Evolution and Mechanical Behavior of High Temperature Solders: Effects of High Temperature Aging

    Science.gov (United States)

    Hasnine, M.; Tolla, B.; Vahora, N.

    2018-04-01

    This paper explores the effects of aging on the mechanical behavior, microstructure evolution and IMC formation on different surface finishes of two high temperature solders, Sn-5 wt.% Ag and Sn-5 wt.% Sb. High temperature aging showed significant degradation of Sn-5 wt.% Ag solder hardness (34%) while aging has little effect on Sn-5 wt.% Sb solder. Sn-5 wt.% Ag experienced rapid grain growth as well as the coarsening of particles during aging. Sn-5 wt.% Sb showed a stable microstructure due to solid solution strengthening and the stable nature of SnSb precipitates. The increase of intermetallic compound (IMC) thickness during aging follows a parabolic relationship with time. Regression analysis (time exponent, n) indicated that IMC growth kinetics is controlled by a diffusion mechanism. The results have important implications in the selection of high temperature solders used in high temperature applications.

  9. Materials Science of High-Level Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

    Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

    2009-01-01

    With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams

  10. Filter-based Aerosol Measurement Experiments using Spherical Aerosol Particles under High Temperature and High Pressure

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Chan; Jung, Woo Young; Lee, Hyun Chul; Lee, Doo Young [FNC TECH., Yongin (Korea, Republic of)

    2016-05-15

    Optical Particle Counter (OPC) is used to provide real-time measurement of aerosol concentration and size distribution. Glass fiber membrane filter also be used to measure average mass concentration. Three tests (MTA-1, 2 and 3) have been conducted to study thermal-hydraulic effect, a filtering tendency at given SiO{sub 2} particles. Based on the experimental results, the experiment will be carried out further with a main carrier gas of steam and different aerosol size. The test results will provide representative behavior of the aerosols under various conditions. The aim of the tests, MTA 1, 2 and 3, are to be able to 1) establish the test manuals for aerosol generation, mixing, sampling and measurement system, which defines aerosol preparation, calibration, operating and evaluation method under high pressure and high temperature 2) develop commercial aerosol test modules applicable to the thermal power plant, environmental industry, automobile exhaust gas, chemical plant, HVAC system including nuclear power plant. Based on the test results, sampled aerosol particles in the filter indicate that important parameters affecting aerosol behavior aerosols are 1) system temperature to keep above a evaporation temperature of ethanol and 2) aerosol losses due to the settling by ethanol liquid droplet.

  11. 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

  12. 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

  13. Method of nuclear reactor control using a variable temperature load dependent set point

    International Nuclear Information System (INIS)

    Kelly, J.J.; Rambo, G.E.

    1982-01-01

    A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

  14. 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

  15. Nuclear hyperfine interactions and chemical bonding in high TC superconductors

    International Nuclear Information System (INIS)

    Danon, J.

    1987-01-01

    Nuclear quadrupole resonances of Cu 63 and Fe 57 Moessbauer spectroscopy of the high temperature superconductor YBa 2 Cu 3 O 7-γ e described together with synchrotron radiation studies of the copper oxidation states in this material. The Moessbauer spectra of 57 Fe in the two distinct crystallographic sites of the Cu atoms in YBa 2 Cu 3 O 7-γ are very similar from the quadrupole coupling point of view although exhibiting markedly different values for the isomer shift. The role of oxygen vacancies in the hyperfine interactions is discussed. (author) [pt

  16. 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)

  17. Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

    KAUST Repository

    Salama, Amgad

    2015-11-01

    One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since the dawn of nuclear era. Albeit the relatively large number of research works that have been conducted to investigate temperature distribution surrounding waste canisters, they all abide to consider the host formations as homogeneous and isotropic. While this could be the case in some subsurface settings, in most cases, this is not true. In other words, subsurface formations are, in most cases, inherently anisotropic and heterogeneous. In this research, we show that even a slight difference in anisotropy of thermal conductivity of host rock with direction could have interesting effects on temperature fields. We investigate the effect of anisotropy angle (the angle the principal direction of anisotropy is making with the coordinate system) on the temperature field as well as on the maximum temperature attained in different barrier systems. This includes 0°, 30°, 45°, 60°, and 90°in addition to the isotropic case as a reference. We also consider the effect of anisotropy ratio (the ratio between the principal direction anisotropies) on the temperature fields and maximum temperature history. This includes ratios ranging between 1.5 and 4. Interesting patterns of temperature fields and profiles are obtained. It is found that the temperature contours are aligned more towards the principal direction of anisotropy. Furthermore the peak temperature in the buffer zone is found to be larger the smaller the anisotropy angle and vice versa. © 2015 Elsevier Ltd. All rights reserved.

  18. Nonlinear effects of high temperature on buckling of structural elements

    International Nuclear Information System (INIS)

    Iyengar, N.G.R.

    1975-01-01

    Structural elements used in nuclear reactors are subjected to high temperatures. Since with increase in temperature there is a gradual fall in the elastic modulus and the stress-strain relationship is nonlinear at these operating load levels, a realistic estimate of the buckling load should include this nonlinearity. In this paper the buckling loads for uniform columns with circular and rectangular cross-sections and different boundary conditions under high temperature environment are estimated. The stress-strain relationship for the material has been assumed to follow inverse Ramberg-Osgood law. In view of the fact that no closed form solutions are possible, approximate methods like perturbation and Galerkin techniques are used. Further, the solution for general value for 'm' is quite involved. Results have been obtained with values for 'm' as 3 and 5. Studies reveal that the influence of material nonlinearity on the buckling load is of the softening type, and it increases with increase in the value of 'm'. The nonlinear effects are more for clamped boundaries than for simply supported boundaries. For the first mode analysis both the methods are powerful. It is, however, felt that for higher modes the Galerkin method might be better in view of its simplicity. This investigation may be considered as a step towards a more general solution

  19. Sandia_HighTemperatureComponentEvaluation_2015

    Energy Technology Data Exchange (ETDEWEB)

    Cashion, Avery T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    The objective of this project is to perform independent evaluation of high temperature components to determine their suitability for use in high temperature geothermal tools. Development of high temperature components has been increasing rapidly due to demand from the high temperature oil and gas exploration and aerospace industries. Many of these new components are at the late prototype or first production stage of development and could benefit from third party evaluation of functionality and lifetime at elevated temperatures. In addition to independent testing of new components, this project recognizes that there is a paucity of commercial-off-the-shelf COTS components rated for geothermal temperatures. As such, high-temperature circuit designers often must dedicate considerable time and resources to determine if a component exists that they may be able to knead performance out of to meet their requirements. This project aids tool developers by characterization of select COTS component performances beyond published temperature specifications. The process for selecting components includes public announcements of project intent (e.g., FedBizOps), direct discussions with candidate manufacturers,and coordination with other DOE funded programs.

  20. A survey of reference electrodes for high temperature waters

    International Nuclear Information System (INIS)

    Molander, A.; Eriksson, Sture; Pein, K.

    2000-11-01

    In nuclear power plants, corrosion potential measurements are used to follow the conditions for different corrosion types in reactor systems, particularly IGSCC in BWRs. The goal of this work has been to give a survey of reference electrodes for high temperature water, both those that are used for nuclear environments and those that are judged to possible future development. The reference electrodes that are used today in nuclear power plants for corrosion potential measurements are of three types. Silver chloride electrodes, membrane electrodes and platinum electrodes (hydrogen electrodes). The principals for their function is described as well as the conversion of measured potentials to the SHE scale (Standard Hydrogen Electrode). Silver chloride electrodes consist of an inner reference system of silver chloride in equilibrium with a chloride solution. The silver chloride electrode is the most common reference electrode and can be used in several different systems. Platinum electrodes are usually more robust and are particularly suitable to use in BWR environment to follow the hydrogen dosage, but have limitations at low and no hydrogen dosage. Ceramic membrane electrodes can be with different types of internal reference system. They were originally developed for pH measurements in high temperature water. If pH is constant, the membrane electrode can be used as reference electrode. A survey of ceramic reference electrodes for high temperature water is given. A ceramic membrane of the type used works as an oxygen conductor, so the potential and pH in surrounding medium is in equilibrium with the internal reference system. A survey of the lately development of electrodes is presented in order to explain why the different types of electrodes are developed as well as to give a background to the possibilities and limitations with the different electrodes. Possibilities of future development of electrodes are also given. For measurements at low or no hydrogen dosage