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Sample records for htgr process heat

  1. Technical review of process heat applications using the HTGR

    International Nuclear Information System (INIS)

    Brierley, G.

    1976-06-01

    The demand for process heat applications is surveyed. Those applications which can be served only by the high temperature gas-cooled reactor (HTGR) are identified and the status of process heat applications in Europe, USA, and Japan in December 1975 is discussed. Technical problems associated with the HTGR for process heat applications are outlined together with an appraisal of the safety considerations involved. (author)

  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. Prospects of HTGR process heat application and role of HTTR

    International Nuclear Information System (INIS)

    Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    At Japan Atomic Energy Research Institute, an effort on development of process heat application with high temperature gas cooled reactor (HTGR) has been continued for providing a future clean alternative to the burning of fossil energy for the production of industrial process heat. The project is named 'HTTR Heat Utilization Project', which includes a demonstration of hydrogen production using the first Japanese HTGR of High Temperature Engineering Test Reactor (HTTR). In the meantime, some countries, such as China, Indonesia, Russia and South Africa are trying to explore the HTGR process heat application for industrial use. One of the key issues for this application is economy. It has been recognized for a long time and still now that the HTGR heat application system is not economically competitive to the current fossil ones, because of the high cost of the HTGR itself. However, the recent movement on the HTGR development, as represented by South Africa Pebble Beds Modular Reactor (SA-PBMR) Project, has revealed that the HTGRs are well economically competitive in electricity production to fossil fuel energy supply under a certain condition. This suggests that the HTGR process heat application will also possibly get economical in the near future. In the present paper, following a brief introduction describing the necessity of the HTGRs for the future process heat application, Japanese activities and prospect of the development on the process heat application with the HTGRs are described in relation with the HTTR Project. In conclusion, the process heat application system with HTGRs is thought technically and economically to be one of the most promising applications to solve the global environmental issues and energy shortage which may happen in the future. However, the commercialization for the hydrogen production system from water, which is the final goal of the HTGR process heat application, must await the technology development to be completed in 2030's at the

  4. Process heat utilization from HTGR type reactors

    International Nuclear Information System (INIS)

    1985-01-01

    Work performed by the Special Research Unit 163 to supplement industrial development projects in the subject field was devoted to specific problems. The major goal was to analyse available industrial developments for potential improvements in terms of process design and engineering in line with the latest know-how, in order to enhance the economic efficiency of available techniques and methods. So research into coal gasification by nuclear processes concentrated on the potentials of a method allowing significantly higher gasification temperatures due to the use of a so-called high-temperature heat pump operating on the basis of the gas turbine principle. Exergetic analyses were made for the processes using nuclear heat in order to optimise their energy consumption. Major steps in these processes are gas purification and gas separation. Especially for the latter step, novel techniques were studied and tested on lab scale, results being used for development towards technical scale application. One novel technique is a method for separating hydrogen from methane and carbon monoxide by means of a gas turbine process step, another research task resulted in a novel absorption technique in the liquid phase. Further, alternative solutions were studied which, other than the conventional gasification processes, comprise electrochemical and other chemical process steps. The important research topic concerned with the kinetics of coal gasification was made part of a special research program on the level of fundamental research. (orig./GL) [de

  5. European research and development on HTGR process heat applications

    International Nuclear Information System (INIS)

    Verfondern, Karl; Lensa, Werner von

    2003-01-01

    The High-Temperature Gas-Cooled Reactor represents a suitable and safe concept of a future nuclear power plant with the potential to produce process heat to be utilized in many industrial processes such as reforming of natural gas, coal gasification and liquefaction, heavy oil recovery to serve for the production of the storable commodities hydrogen or energy alcohols as future transportation fuels. The paper will include a description of the broad range of applications for HTGR process heat and describe the results of the German long-term projects ''Prototype Nuclear Process Heat Reactor Project'' (PNP), in which the technical feasibility of an HTGR in combination with a chemical facility for coal gasification processes has been proven, and ''Nuclear Long-Distance Energy Transportation'' (NFE), which was the demonstration and verification of the closed-cycle, long-distance energy transmission system EVA/ADAM. Furthermore, new European research initiatives are shortly described. A particular concern is the safety of a combined nuclear/chemical facility requiring a concept against potential fire and explosion hazards. (author)

  6. Design of the HTGR for process heat applications

    International Nuclear Information System (INIS)

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

    1980-05-01

    This paper discusses a design study of an advanced 842-MW(t) HTGR with a reactor outlet temperature of 850 0 C (1562 0 F), coupled with a chemical process whose product is hydrogen (or a mixture of hydrogen and carbon monoxide) generated by steam reforming of a light hydrocarbon mixture. This paper discusses the plant layout and design for the major components of the primary and secondary heat transfer systems. Typical parametric system study results illustrate the capability of a computer code developed to model the plant performance and economics

  7. Nuclear heat source design for an advanced HTGR process heat plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; O'Hanlon, T.W.

    1983-01-01

    A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

  8. HTGR high temperature process heat design and cost status report

    International Nuclear Information System (INIS)

    1981-12-01

    This report describes the status of the studies conducted on the 850 0 C ROT indirect cycle and the 950 0 C ROT direct cycle through the end of Fiscal Year 1981. Volume I provides summaries of the design and optimization studies and the resulting capital and product costs, for the HTGR/thermochemical pipeline concept. Additionally, preliminary evaluations are presented for coupling of candidate process applications to the HTGR system

  9. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  10. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  11. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

  12. HTGR high temperature process heat design and cost status report. Volume II. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-12-01

    Information is presented concerning the 850/sup 0/C IDC reactor vessel; primary cooling system; secondary helium system; steam generator; heat cycle evaluations for the 850/sup 0/C IDC plant; 950/sup 0/C DC reactor vessel; 950/sup 0/C DC steam generator; direct and indirect cycle reformers; methanation plant; thermochemical pipeline; methodology for screening candidate synfuel processes; ECCG process; project technical requirements; process gas explosion assessment; HTGR program economic guidelines; and vendor respones.

  13. HTGR high temperature process heat design and cost status report. Volume II. Appendices

    International Nuclear Information System (INIS)

    1981-12-01

    Information is presented concerning the 850 0 C IDC reactor vessel; primary cooling system; secondary helium system; steam generator; heat cycle evaluations for the 850 0 C IDC plant; 950 0 C DC reactor vessel; 950 0 C DC steam generator; direct and indirect cycle reformers; methanation plant; thermochemical pipeline; methodology for screening candidate synfuel processes; ECCG process; project technical requirements; process gas explosion assessment; HTGR program economic guidelines; and vendor respones

  14. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

    The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

  15. 60-MW/sub t/ methanation plant design for HTGR process heat

    International Nuclear Information System (INIS)

    Davis, C.R.; Arcilla, N.T.; Hui, M.M.; Hutchins, B.A.

    1982-07-01

    This report describes a 60 MW(t) Methanation Plant for generating steam for industrial applications. The plant consists of four 15 MW(t) methanation trains. Each train is connected to a pipeline and receives synthesis gas (syngas) from a High Temperature Gas-Cooled Reactor Reforming (HTGR-R) plant. Conversion of the syngas to methane and water releases exothermic heat which is used to generate steam. Syngas is received at the Methanation Plant at a temperature of 80 0 F and 900 psia. One adiabatic catalytic reactor and one isothermal catalytic reactor, in each methanation train, converts the syngas to 92.2% (dry bases) methane. Methane and condensate are returned at temperatures of 100 to 125 0 F and at pressures of 860 to 870 psia to the HTGR-R plant for the reproduction of syngas

  16. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

    1981-01-15

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

  17. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    International Nuclear Information System (INIS)

    Fritsching, G.; Wolf, G.

    1981-01-01

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

  18. Reduced risk HTGR concept for industrial heat application

    International Nuclear Information System (INIS)

    Boardman, C.E.; Lipps, A.J.

    1982-01-01

    The industrial process heat market has been identified as major market for the High Temperature Gas-Cooled Reactor (HTGR), however, this market introduces stringent availability requirements on the reactor system relative to electric plants which feed a large existing grid. The characteristics and requirements of the industrial heat markets are summarized; the risks associated with serving this market with a single large HTGR will be discussed; and the modular concept, which has the potential to reduce both safety and investment risks, will be described. The reference modular concept described consists of several small, relatively benign nuclear heat sources linked together to supply heat energy to a balance-of-plant incorporating a process gas train/thermochemical pipe line system and a normal steam-electric plant

  19. HTGR process heat program design and analysis. Semiannual progress report, October 1, 1979-March 28, 1980

    International Nuclear Information System (INIS)

    1980-10-01

    This report summarizes the results of concept design studies implemented at General Atomic Company (GA) during the first half of FY-80. The studies relate to a plant design for an 842-MW(t) High-Temperature Gas-Cooled Reactor utilizing an intermediate helium heat transfer loop to provide high temperature thermal energy for the production of hydrogen or synthesis gas (H 2 + CO) by steam-reforming a light hydrocarbon. Basic carbon sources may be coal, residual oil, or oil shale. Work tasks conducted during this period included the 842-MW(t) plant concept design and cost estimate for an 850 0 C reactor outlet temperature. An assessment of the main-loop cooling shutdown system is reported. Major component cost models were prepared and programmed into the Process Heat Reactor Evaluation and Design (PHRED) code

  20. Screening of synfuel processes for HTGR application

    International Nuclear Information System (INIS)

    1981-02-01

    The aim of this study is to select for further study, the several synfuel processes which are the most attractive for application of HTGR heat and energy. In pursuing this objective, the Working Group identified 34 candidate synfuel processes, cut the number of processes to 16 in an initial screening, established 11 prime criteria with weighting factors for use in screening the remaining processes, developed a screening methodology and assumptions, collected process energy requirement information, and performed a comparative rating of the processes. As a result of this, three oil shale retorting processes, two coal liquefaction processes and one coal gasification process were selected as those of most interest for further study at this time

  1. HTGR process heat program design and analysis. Final report, FY-79

    International Nuclear Information System (INIS)

    1979-12-01

    This report summarizes the results of concept design studies at General Atomic Company during FY-79 for an 842-MW(t) Very High Temperature Reactor (VHTR) utilizing an intermediate helium heat transfer loop to provide thermal energy for the production of hydrogen or reducing gas (H 2 + CO) by steam-reforming of a light hydrocarbon. Basic carbon sources may be coal, residual oil, or oil shale. The report summarizes conceptual design tasks conducted on the prestressed concrete reactor vessel, thermal barrier, intermediate heat exchanger, reformer, and steam generator. The substantial completion of first generation programming for a performance/optimization code and the preparation of a topical safety report and other safety evaluation studies are reported. The completion of balance of plant criteria specifications and a balance of plant cost estimate is also reported

  2. Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

    International Nuclear Information System (INIS)

    McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

    1980-02-01

    Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed

  3. Overview of HTGR heat utilization system development at JAERI

    International Nuclear Information System (INIS)

    Miyamoto, Y.; Shiozawa, S.; Ogawa, M.; Akino, N.; Shimizu, S.; Hada, K.; Inagaki, Y.; Onuki, K.; Takeda, T.; Nishihara, T.

    1998-01-01

    The Japan Atomic Energy Research Institute (JAERI) has conducted research and development of nuclear heat utilization systems of a High Temperature Gas cooled Reactor (HTGR), which are capable to meet a large amount of energy demand without significant CO 2 emission to relax the global warming issue. The High Temperature engineering Test Reactor (HTTR) with thermal output of 30 MW and outlet coolant temperature of 950 deg C, the first HTGR in Japan, is under construction on the JAERI site, and its first criticality is scheduled for mid-1998. After the reactor performance and safety demonstration tests for several years, a hydrogen production system will be connected to the HTTR. A demonstration program on hydrogen production started in January 1997, in JAERI, as a study consigned by the Science and Technology Agency. A hydrogen production system connected to the HTTR is designed to be able to produce hydrogen by steam reforming of natural gas, using nuclear heat of 10 MW from the HTTR. The safety principle and standard are investigated for the HTTR hydrogen production system. In order to confirm safety, controllability and performance of key components in the HTTR hydrogen production system, an out-of-pile test facility on the scale of approximately 1/30 of the HTTR hydrogen production system is installed. It is equipped with an electric heater as a heat source instead of the HTTR. The out-of-pile test will be performed for four years after 2001. The HTTR hydrogen production system will be demonstratively operated after 2005 at its earliest plan. Other basic studies on the hydrogen production system using thermochemical water splitting, an iodine sulphur (IS) process, and technology of distant heat transport with microencapsulated phase change material have been carried out for more effective and various uses of nuclear heat. (author)

  4. HTGR-GT closed-cycle gas turbine: a plant concept with inherent cogeneration (power plus heat production) capability

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1980-04-01

    The high-grade sensible heat rejection characteristic of the high-temperature gas-cooled reactor-gas turbine (HTGR-GT) plant is ideally suited to cogeneration. Cogeneration in this nuclear closed-cycle plant could include (1) bottoming Rankine cycle, (2) hot water or process steam production, (3) desalination, and (4) urban and industrial district heating. This paper discusses the HTGR-GT plant thermodynamic cycles, design features, and potential applications for the cogeneration operation modes. This paper concludes that the HTGR-GT plant, which can potentially approach a 50% overall efficiency in a combined cycle mode, can significantly aid national energy goals, particularly resource conservation

  5. Development of structural design procedure of plate-fin heat exchanger for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Mizokami, Yorikata, E-mail: yorikata_mizokami@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-Chome, Hyogo-ku, Kobe 652-8585 (Japan); Igari, Toshihide [Mitsubishi Heavy Industries, Ltd., 5-717-1, Fukahori-machi, Nagasaki 851-0392 (Japan); Kawashima, Fumiko [Kumamoto University, 39-1 Kurokami 2-Chome, Kumamoto 860-8555 (Japan); Sakakibara, Noriyuki [Mitsubishi Heavy Industries, Ltd., 5-717-1, Fukahori-machi, Nagasaki 851-0392 (Japan); Tanihira, Masanori [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo 108-8215 (Japan); Yuhara, Tetsuo [The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Hiroe, Tetsuyuki [Kumamoto University, 39-1 Kurokami 2-Chome, Kumamoto 860-8555 (Japan)

    2013-02-15

    Highlights: ► We propose high temperature structural design procedure for plate-fin heat exchanger ► Allowable stresses for brazed structures will be newly discussed ► Validity of design procedure is confirmed by carrying out partial model tests ► Proposed design procedure is applied to heat exchangers for HTGR. -- Abstract: Highly efficient plate-fin heat exchanger for application to HTGR has been focused on recently. Since this heat exchanger is fabricated by brazing a lot of plates and fins, a new procedure for structural design of brazed structures in the HTGR temperature region up to 950 °C is required. Firstly in this paper influences on material strength due to both thermal aging during brazing process and helium gas environment were experimentally examined, and failure mode and failure limit of brazed side-bar structures were experimentally clarified. Secondly allowable stresses for aging materials and brazed structures were newly determined on the basis of the experimental results. For the purpose of validating the structural design procedure including homogenization FEM modeling, a pressure burst test and a thermal fatigue test of partial model for plate-fin heat exchanger were carried out. Finally, results of reference design of plate-fin heat exchangers of recuperator and intermediate heat exchanger for HTGR plant were evaluated by the proposed design criteria.

  6. Present status of HTGR projects and their heat applications in Russia

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Glushkov, E.S.; Kukharkin, N.E.; Ponomarev-Stepnoi, N.N.

    1996-01-01

    This paper describes the main technical decision and parameters of the HTGR of different power and considers a few schemes of HTGR plants with a gas turbine cycle. Also, the future prospects on heat utilization of HTGR in Russia is presented. (J.P.N.)

  7. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

  8. 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

  9. Safety analysis of coupling system of hybrid (MED-RO) nuclear desalination system utilising waste heat from HTGR

    International Nuclear Information System (INIS)

    Raha, Abhijit; Kishore, G.; Rao, I.S.; Adak, A.K.; Srivastava, V.K.; Prabhakar, S.; Tewari, P.K.

    2010-01-01

    To meet the generation IV goals, High Temperature Gas Cooled Reactors (HTGRs) are designed to have relatively higher thermal efficiency and enhanced safety and environmental characteristics. It can provide energy for combined production of hydrogen, electricity and other industrial applications. The waste heat available in the HTGR power cycle can also be utilized for the desalination of seawater for producing potable water. Desalination is an energy intensive process, so use of waste heat from HTGR certainly makes desalination process more affordable to create fresh water resources. So design of the coupling system, as per the safety design requirement of nuclear desalination plant, of desalination plant with HTGR is very crucial. In the first part of this paper, design of the coupling system between hybrid Multi Effect Desalination-Reverse Osmosis (MED-RO) nuclear desalination plant and HTGR to utilize the waste heat in HTGR are discussed. In the next part deterministic safety analysis of the designed coupling system of are presented in detail. It was found that all the coupling system meets the acceptance criteria for all the Postulated Initiating Events (PIE's) limited to DBA. (author)

  10. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  11. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  12. The effect of creep-fatigue damage relationships upon HTGR heat exchanger design

    International Nuclear Information System (INIS)

    Kozina, M.M.; King, J.H.; Basol, M.

    1984-01-01

    Materials for heat exchangers in the high temperature gas-cooled reactor (HTGR) are subjected to cyclic loading, extending the necessity to design against fatigue failure into the temperature region where creep processes become significant. Therefore, the fatigue life must be considered in terms of creep-fatigue interaction. In addition, since HTGR heat exchangers are subjected to holds at constant strain levels or constant stress levels in high-temperature environments, the cyclic life is substantially reduced. Of major concern in the design and analysis of HTGR heat exchangers is the accounting for the interaction of creep and fatigue. The accounting is done in conformance to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Case N-47, which allows the use of the linear damage criterion for interaction of creep and fatigue. This method separates the damage incurred in the material into two parts: one due to fatigue and one due to creep. The summation of the creep-fatigue damage must be less than 1.0. Recent material test data have indicated that the assumption of creep and fatigue damage equals unity at failure may not always be valid for materials like Alloy 800H, which is used in the higher temperature sections of HTGR steam generators. Therefore, a more conservative creep-fatigue damage relationship was postulated for Alloy 800H. This more conservative bilinear damage relationship consists of a design locus drawn from D F =1.0, D C =0 to D F =0.1, D C =0.1 to D F =0, D C =1.0. D F is the fatigue damage and D C is the creep damage. A more conservative damage relationship for 2-1/4 Cr-1 Mo material consisted of including factors that degrade the fatigue curves. These revised relationships were used in a structural evaluation of the HTGR steam cycle/cogeneration (SC/C) steam generator design. The HTGR-SC/C steam generator, a once-through type, is comprised of an economizer-evaporator-superheater (ESS) helical bundle of 2-1/4 Cr-1

  13. Promising materials for HTGR high temperature heat exchangers

    International Nuclear Information System (INIS)

    Kuznetsov, E.V.; Tokareva, T.B.; Ryabchenkov, A.V.; Novichkova, O.V.; Starostin, Yu.D.

    1989-01-01

    The service conditions for high-temperature heat-exchangers with helium coolant of HTGRs and requirements imposed on materials for their production are discussed. The choice of nickel-base alloys with solid-solution hardening for long-term service at high temperatures is grounded. Results of study on properties and structure of types Ni-25Cr-5W-5Mo and Ni-20Cr-20W alloy in the temperature range of 900 deg. - 1,000 deg. C are given. The ageing of Ni-25Cr-5W-5Mo alloy at 900 deg. - 950 deg. C results in decreased corrosion-mechanical properties and is caused by the change of structural metal stability. Alloy with 20% tungsten retains a high stability of both structure and properties after prolonged exposure in helium at above temperatures. The alloy has also increased resistance to delayed fracture and low-cycle fatigue at high temperatures. The developed alloy of type Ni-20Cr-20W with microalloying is recommended for production of tubes for HTGR high-temperature heat-exchangers with helium coolant. (author). 3 refs, 8 figs

  14. Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

    International Nuclear Information System (INIS)

    Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

    1977-04-01

    The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

  15. Process control of an HTGR fuel reprocessing cold pilot plant

    International Nuclear Information System (INIS)

    Rode, J.S.

    1976-10-01

    Development of engineering-scale systems for a large-scale HTGR fuel reprocessing demonstration facility is currently underway in a cold pilot plant. These systems include two fluidized-bed burners, which remove the graphite (carbon) matrix from the crushed HTGR fuel by high temperature (900 0 C) oxidation. The burners are controlled by a digital process controller with an all analog input/output interface which has been in use since March, 1976. The advantages of such a control system to a pilot plant operation can be summarized as follows: (1) Control loop functions and configurations can be changed easily; (2) control constants, alarm limits, output limits, and scaling constants can be changed easily; (3) calculation of data and/or interface with a computerized information retrieval system during operation are available; (4) diagnosis of process control problems is facilitated; and (5) control panel/room space is saved

  16. Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

    The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

  17. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  18. HTGR generic technology program. Semiannual report ending March 31, 1980

    International Nuclear Information System (INIS)

    1980-05-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an MEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbine and process heat plants

  19. Component design considerations for gas turbine HTGR waste-heat power plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Vrable, D.L.

    1976-01-01

    Component design considerations are described for the ammonia waste-heat power conversion system of a large helium gas-turbine nuclear power plant under development by General Atomic Company. Initial component design work was done for a reference plant with a 3000-MW(t) High-Temperature Gas-Cooled Reactor (HTGR), and this is discussed. Advanced designs now being evaluated include higher core outlet temperature, higher peak system pressures, improved loop configurations, and twin 4000-MW(t) reactor units. Presented are the design considerations of the major components (turbine, condenser, heat input exchanger, and pump) for a supercritical ammonia Rankine waste heat power plant. The combined cycle (nuclear gas turbine and waste-heated plant) has a projected net plant efficiency of over 50 percent. While specifically directed towards a nuclear closed-cycle helium gas-turbine power plant (GT-HTGR), it is postulated that the bottoming waste-heat cycle component design considerations presented could apply to other low-grade-temperature power conversion systems such as geothermal plants

  20. Design and thermal dynamic analyses on the intermediate heat exchanger for HTGR

    International Nuclear Information System (INIS)

    Mori, M.; Mizuno, M.; Ito, M.; Urabe, S.

    1986-01-01

    The intermediate heat exchanger (IHX), one of the most important components in the high temperature gas cooled reactor (HTGR), is a high performance helium/helium (He/He) heat exchanger operated at a very high temperature above 900 0 C to transmit the nuclear heat from the reactor core to the nuclear heat utilization systems such as the chemical plant. Having to meet, in addition, the requirement of the pressure boundary as the Class-1 it demands the accurate estimation of thermal performance and analytical prediction of thermal behaviors to secure its integrity throughout the service life. In the present works, the newly-developed analytical codes carry out designing thermal performance and analyzing dynamic thermal behaviors of the IHX. These codes have been developed on a great deal of data and studies related to the research and development on the 1.5 MWt- and the 25 MWt-IHXs. This paper shows the design on the IHX, the results of the dynamic analyses on the 1.5 MWt-IHX with the comparison to the experimental data and the analytical predictions of the dynamic thermal behaviors on the 25 MWt-IHX. The results calculated are in fairly good agreement with the experimental data on the 1.5 MWt-IHX, the fact that has verified the analytical codes to be reasonable and much useful for the thermal design of the IHX. These presented results and data are available for the design of the IHX of HTGR

  1. Development of processes and equipment for the refabrication of HTGR fuels

    International Nuclear Information System (INIS)

    Sease, J.D.; Lotts, A.L.

    1976-06-01

    Refabrication is in the step in the HTGR thorium fuel cycle that begins with a nitrate solution containing 238 U and culminates in the assembly of this material into fuel elements for use in an HTGR. Refabrication of HTGR fuel is essentially a manufacturing operation and consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and SiC, preparation of fuel rods, and assembly of fuel rods in fuel elements. All the equipment for refabrication of 238 U-containing fuel must be designed for completely remote operation and maintenance in hot cell facilities. This paper describes the status of processes and equipment development for the remote refabrication of HTGR fuels. The feasibility of HTGR refabrication processes has been proven by laboratory development. Engineering-scale development is now being performed on a unit basis on the majority of the major equipment items. Engineering-scale equipment described includes full-scale resin loading equipment, a 5-in.-dia (0.13-m) microsphere coating furnace, a fuel rod forming machine, and a cure-in-place furnace

  2. Liquid-metal-gas heat exchanger for HTGR type reactors

    International Nuclear Information System (INIS)

    Werth, G.

    1980-01-01

    The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL) [de

  3. Safety concerns and suggested design approaches to the HTGR Reformer process concept

    International Nuclear Information System (INIS)

    Green, R.C.

    1981-09-01

    This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept

  4. Safety concerns and suggested design approaches to the HTGR Reformer process concept

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.C.

    1981-09-01

    This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept.

  5. Numerical simulation of flow field in cooling tower of passive residual heat removal system of HTGR

    International Nuclear Information System (INIS)

    Li Xiaowei; Zhang Li; Wu Xinxin; He Shuyan

    2011-01-01

    Environmental wind will influence the working conditions of natural convection cooling tower. The velocity and temperature fields in the natural convection cooling tower of the HTGR residual heat removal system at different environmental wind velocities were numerically simulated. The results show that, if there is no wind baffle, the flow in the cooling tower is blocked when environmental wind velocity is higher than 6 m/s, residual heat can hardly be removed, and when wind velocity is higher than 9 m/s, the air even flow downwards in the tower, so wind baffle is very necessary. With the wind baffle installed, the cooling tower works well at the wind speed even higher than 9 m/s. The optimum baffle size and positions are also analyzed. (authors)

  6. Multiregional coupled conduction--convection model for heat transfer in an HTGR core

    International Nuclear Information System (INIS)

    Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

    1978-01-01

    HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

  7. Passive afterheat removal in the HTGR with the liner cooling system as a heat sink

    International Nuclear Information System (INIS)

    Rehm, W.; Jahn, W.; Verfondern, K.

    1984-09-01

    The report deals with the transients of temperature and system pressure and the fission product behaviour in the primary circuit of an HTGR during passive afterheat removal, where the liner cooling system of the PCRV serves as a heat sink. The analysis has been made for the PNP-500-reactor representing nuclear plants with medium thermal power. The investigations show that the liner cooling system is able to control a core heatup. High temperature loads are encountered in the upper core region. In the case of a reactor under pressure the fuel elements and the primary circuit remain intact as the first and second barriers for fission products. In the case of a depressurized primary circuit the liner cooling system also keeps the PCRV at normal operating temperatures. The effects of a core heatup on component damage and release of fission products are thus limited. (orig.) [de

  8. Assessment of the SRI Gasification Process for Syngas Generation with HTGR Integration -- White Paper

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-04-01

    This white paper is intended to compare the technical and economic feasibility of syngas generation using the SRI gasification process coupled to several high-temperature gas-cooled reactors (HTGRs) with more traditional HTGR-integrated syngas generation techniques, including: (1) Gasification with high-temperature steam electrolysis (HTSE); (2) Steam methane reforming (SMR); and (3) Gasification with SMR with and without CO2 sequestration.

  9. HTGR depressurization analysis

    International Nuclear Information System (INIS)

    Boccio, J.L.; Colman, J.; Skalyo, J.; Beerman, J.

    1979-01-01

    Relaxation of the prima facie assumption of complete mixing of primary and secondary containment gases during HTGR depressurization has led to a study program designed to identify and selectively quantify the relevant gas dynamic processes which prevail during the depressurization event. Uncertainty in the degree of gas mixedness naturally leads to uncertainty in containment vessel design pressure and heat loads and possible combustion hazards therein. This paper succinctly details an analytical approach and modeling methodology of the exhaust jet structure/containment vessel interaction during penetration failures. (author)

  10. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1980

    International Nuclear Information System (INIS)

    1980-11-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an LEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbines and process heat plants

  11. Life time test of a partial model of HTGR helium-helium heat exchanger

    International Nuclear Information System (INIS)

    Kitagawa, Masaki; Hattori, Hiroshi; Ohtomo, Akira; Teramae, Tetsuo; Hamanaka, Junichi; Itoh, Mitsuyoshi; Urabe, Shigemi

    1984-01-01

    Authors had proposed a design guide for the HTGR components and applied it to the design and construction of the 1.5 Mwt helium heat exchanger test loop for the nuclear steel making under the financial support of the Japanese Ministry of International Trade and Industry. In order to assure that the design method covers all the conceivable failure mode and has enough safety margin, a series of life time tests of partial model may be needed. For this project, three types of model tests were performed. A life time test of a partial model of the center manifold pipe and eight heat exchanger tubes were described in this report. A damage criterion with a set of material constants and a simplified method for stress-strain analysis for stub tube under three dimensional load were newly developed and used to predict the lives of each tube. The predicted lives were compared with the experimental lives and good agreement was found between the two. The life time test model was evaluated according to the proposed design guide and it was found that the guide has a safety factor of approximately 200 in life for this particular model. (author)

  12. Industrial process heat market assessment

    International Nuclear Information System (INIS)

    Bresnick, S.

    1981-12-01

    This report is designed to be a reference resource, giving a broad perspective of the potential HTGR market for industrial process heat. It is intended to serve as a briefing document for those wishing to obtain background information and also to serve as a starting point from which more detailed and refined studies may be undertaken. In doing so, the report presents a qualitative and quantitative description of the industrial process heat market in the US, provides a summary discussion of cogeneration experience to date, and outlines the existing institutional and financial framework for cogeneration. The intent is to give the reader an understanding of the current situation and experience in this area. The cogeneration area in particular is an evolving one because of regulations and tax laws, which are still in the process of being developed and interpreted. The report presents the latest developments in regulatory and legislative activities which are associated with that technology. Finally, the report presents a brief description of the three HTGR systems under study during the current fiscal year and describes the specific market characteristics which each application is designed to serve

  13. Industrial process heat market assessment

    Energy Technology Data Exchange (ETDEWEB)

    Bresnick, S.

    1981-12-01

    This report is designed to be a reference resource, giving a broad perspective of the potential HTGR market for industrial process heat. It is intended to serve as a briefing document for those wishing to obtain background information and also to serve as a starting point from which more detailed and refined studies may be undertaken. In doing so, the report presents a qualitative and quantitative description of the industrial process heat market in the US, provides a summary discussion of cogeneration experience to date, and outlines the existing institutional and financial framework for cogeneration. The intent is to give the reader an understanding of the current situation and experience in this area. The cogeneration area in particular is an evolving one because of regulations and tax laws, which are still in the process of being developed and interpreted. The report presents the latest developments in regulatory and legislative activities which are associated with that technology. Finally, the report presents a brief description of the three HTGR systems under study during the current fiscal year and describes the specific market characteristics which each application is designed to serve.

  14. Process behavior and environmental assessment of 14C releases from an HTGR fuel reprocessing facility

    International Nuclear Information System (INIS)

    Snider, J.W.; Kaye, S.V.

    1976-01-01

    Large quantities of 14 CO 2 will be evolved when graphite fuel blocks are burned during reprocessing of spent fuel from HTGR reactors. The possible release of some or all of this 14 C to the environment is a matter of concern which is investigated in this paper. Various alternatives are considered in this study for decontaminating and releasing the process off-gas to the environment. Concomitant radiological analyses have been done for the waste process scenarios to supply the necessary feedbacks for process design

  15. HTGR generic technology program plan (FY 80)

    International Nuclear Information System (INIS)

    1980-01-01

    Purpose of the program is to develop base technology and to perform design and development common to the HTGR Steam Cycle, Gas Turbine, and Process Heat Plants. The generic technology program breaks into the base technology, generic component, pebble-bed study, technology transfer, and fresh fuel programs

  16. Computational model and performance optimization methodology of a compact design heat exchanger used as an IHX in HTGR

    International Nuclear Information System (INIS)

    De la Torre V, R.; Francois L, J. L.

    2017-09-01

    The intermediate heat exchangers (IHX) present in high-temperature gas-cooled reactor (HTGR) present complex operating conditions, characterized by temperature values higher than 1073 K. Conventional designs of tubes and shell have shown disadvantages with respect to compact designs. In this work, computational models of a compact heat exchanger design, the printed circuit, were built under IHX conditions in a HTGR installation. In these models, a detailed geometry was considered in three dimensions, corresponding to a transfer unit of the heat exchanger. Computational fluid dynamics techniques and finite element methods were used to study the thermo-hydraulic and mechanical functioning of the equipment, respectively. The properties of the materials were defined as temperature functions. The thermo-hydraulic results obtained were established as operating conditions in the structural calculations. A methodology was developed based on the analysis of capital and operating costs, which takes into account the heat transfer, pressure drop and the mechanical behavior of the structure, in a single optimization variable. By analyzing the experimental results of other authors, a relationship was obtained between the operation time of the equipment and the maximum effort in the structure, which was used in the model. The results show that the model that allows a greater thermal efficiency differs from the one that has lower total cost per year. (Author)

  17. Process options and projected mass flows for the HTGR refabrication scrap recovery system

    International Nuclear Information System (INIS)

    Tiegs, S.M.

    1979-03-01

    The two major uranium recovery processing options reviewed are (1) internal recovery of the scrap by the refabrication system and (2) transfer to and external recovery of the scrap by the head end of the reprocessing system. Each option was reviewed with respect to equipment requirements, preparatory processing, and material accountability. Because there may be a high cost factor on transfer of scrap fuel material to the reprocessing system for recovery, all of the scrap streams will be recycled internally within the refabrication system, with the exception of reject fuel elements, which will be transferred to the head end of the reprocessing system for uranium recovery. The refabrication facility will be fully remote; thus, simple recovery techniques were selected as the reference processes for scrap recovery. Crushing, burning, and leaching methods will be used to recover uranium from the HTGR refabrication scrap fuel forms, which include particles without silicon carbide coatings, particles with silicon carbide coatings, uncarbonized fuel rods, carbon furnace parts, perchloroethylene distillation bottoms, and analytical sample remnants. Mass flows through the reference scrap recovery system were calculated for the HTGR reference recycle facility operating with the highly enriched uranium fuel cycle. Output per day from the refabrication scrap recovery system is estimated to be 4.02 kg of 2355 U and 10.85 kg of 233 U. Maximum equipment capacities were determined, and future work will be directed toward the development and costing of the scrap recovery system chosen as reference

  18. Design of performance and analysis of dynamic and transient thermal behaviors on the intermediate heat exchanger for HTGR

    International Nuclear Information System (INIS)

    Mori, Michitsugu; Mizuno, Minoru; Itoh, Mitsuyoshi; Urabe, Shigemi

    1985-01-01

    The intermediate heat exchanger (IHX) is designed as the high temperature heat exchanger for HTGR (High Temperature Gas-cooled Reactor), which transmits the primary coolant helium's heat raised up to about 950 0 C in the reactor core to the secondary helium or the nuclear heat utilization. Having to meet, in addition, the requirement of the primary coolant pressure boundary as the Class-1 component, it must be secured integrity throughout the service life. This paper will show (1) the design of the thermal performance; (2) the results of the dynamic analyses of the 1.5 MWt-IHX with its comparison to the experimental data; (3) the analytical predictions of the dynamic thermal behaviors under start-up and of the transient thermal behaviors during the accident on the 25 MWt-IHX. (author)

  19. Acid in perchloroethylene scrubber solutions used in HTGR fuel preparation processes. Analytical chemistry studies

    International Nuclear Information System (INIS)

    Lee, D.A.

    1979-02-01

    Acids and corrosion products in used perchloroethylene scrubber solutions collected from HTGR fuel preparation processes have been analyzed by several analytical methods to determine the source and possible remedy of the corrosion caused by these solutions. Hydrochloric acid was found to be concentrated on the carbon particles suspended in perchloroethylene. Filtration of carbon from the scrubber solutions removed the acid corrosion source in the process equipment. Corrosion products chemisorbed on the carbon particles were identified. Filtered perchloroethylene from used scrubber solutions contained practically no acid. It is recommended that carbon particles be separated from the scrubber solutions immediately after the scrubbing process to remove the source of acid and that an inhibitor be used to prevent the hydrolysis of perchloroethylene and the formation of acids

  20. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  1. Volume 1. Probabilistic analysis of HTGR application studies. Technical discussion

    International Nuclear Information System (INIS)

    May, J.; Perry, L.

    1980-01-01

    The HTGR Program encompasses a number of decisions facing both industry and government which are being evaluated under the HTGR application studies being conducted by the GCRA. This report is in support of these application studies, specifically by developing comparative probabilistic energy costs of the alternative HTGR plant types under study at this time and of competitive PWR and coal-fired plants. Management decision analytic methodology was used as the basis for the development of the comparative probabilistic data. This study covers the probabilistic comparison of various HTGR plant types at a commercial development stage with comparative PWR and coal-fired plants. Subsequent studies are needed to address the sequencing of HTGR plants from the lead plant to the commercial plants and to integrate the R and D program into the plant construction sequence. The probabilistic results cover the comparison of the 15-year levelized energy costs for commercial plants, all with 1995 startup dates. For comparison with the HTGR plants, PWR and fossil-fired plants have been included in the probabilistic analysis, both as steam electric plants and as combined steam electric and process heat plants

  2. NGNP Process Heat Applications: Hydrogen Production Accomplishments for FY2010

    Energy Technology Data Exchange (ETDEWEB)

    Charles V Park

    2011-01-01

    This report summarizes FY10 accomplishments of the Next Generation Nuclear Plant (NGNP) Engineering Process Heat Applications group in support of hydrogen production technology development. This organization is responsible for systems needed to transfer high temperature heat from a high temperature gas-cooled reactor (HTGR) reactor (being developed by the INL NGNP Project) to electric power generation and to potential industrial applications including the production of hydrogen.

  3. Computational model and performance optimization methodology of a compact design heat exchanger used as an IHX in HTGR; Modelo computacional y metodologia de optimizacion del funcionamiento de un intercambiador de calor de diseno compacto empleado como IHX en HTGR

    Energy Technology Data Exchange (ETDEWEB)

    De la Torre V, R.; Francois L, J. L., E-mail: delatorrevaldes@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico)

    2017-09-15

    The intermediate heat exchangers (IHX) present in high-temperature gas-cooled reactor (HTGR) present complex operating conditions, characterized by temperature values higher than 1073 K. Conventional designs of tubes and shell have shown disadvantages with respect to compact designs. In this work, computational models of a compact heat exchanger design, the printed circuit, were built under IHX conditions in a HTGR installation. In these models, a detailed geometry was considered in three dimensions, corresponding to a transfer unit of the heat exchanger. Computational fluid dynamics techniques and finite element methods were used to study the thermo-hydraulic and mechanical functioning of the equipment, respectively. The properties of the materials were defined as temperature functions. The thermo-hydraulic results obtained were established as operating conditions in the structural calculations. A methodology was developed based on the analysis of capital and operating costs, which takes into account the heat transfer, pressure drop and the mechanical behavior of the structure, in a single optimization variable. By analyzing the experimental results of other authors, a relationship was obtained between the operation time of the equipment and the maximum effort in the structure, which was used in the model. The results show that the model that allows a greater thermal efficiency differs from the one that has lower total cost per year. (Author)

  4. 1170-MW(t) HTGR-PS/C plant application study report: SRC-II process application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    The solvent refined coal (SRC-II) process is an advanced process being developed by Gulf Mineral Resources Ltd. (a Gulf Oil Corporation subsidiary) to produce a clean, non-polluting liquid fuel from high-sulfur bituminous coals. The SRC-II commercial plant will process about 24,300 tonnes (26,800 tons) of feed coal per stream day, producing primarily fuel oil plus secondary fuel gases. This summary report describes the integration of a high-temperature gas-cooled reactor operating in a process steam/cogeneration mode (HTGR-PS/C) to provide the energy requirements for the SRC-II process. The HTGR-PS/C plant was developed by General Atomic Company (GA) specifically for industries which require energy in the form of both steam and electricity. General Atomic has developed an 1170-MW(t) HTGR-PS/C design which is particularly well suited to industrial applications and is expected to have excellent cost benefits over other sources of energy

  5. FY1983 HTGR summary level program plan

    International Nuclear Information System (INIS)

    1983-01-01

    The major focus and priority of the FY1983 HTGR Program is the development of the HTGR-SC/C Lead Project through one of the candidate lead utilities. Accordingly, high priority will be given to work described in WBS 04 for site and user specific studies toward the development of the Lead Project. Asessment of advanced HTGR systems will continue during FY1983 in accordance with the High Temperature Process Heat (HTPH) Concept Evaluation Plan. Within the context of that plan, the assessment of the monolithic HTPH concepts has been essentially completed in FY1982 and FY1983 activities and will be limited to documentation only. the major advanced HTGR systems efforts in FY1983 will be focused on the further definition of the Modular Reactor Systems concepts in both the reforming (MRS-R) and Steam Cycle/Cogeneration 9MRS-SC/C) configurations in WBS 41. The effort will concentrate upon key technical issues and trade studies oriented to reduction in expected cost and schedule duration. With regard to the latter, the most significant will be trade study addressing the degree of modularization of reactor plant structures. particular attention will be given to the confinement building which currently defines the critical path for construction

  6. HTGR experience, programs, and future applications

    International Nuclear Information System (INIS)

    Moore, R.A.; Kantor, M.E.; Brey, H.L.; Olson, H.G.

    1982-01-01

    This paper reviews the current status of the programs for the development of high-temperature gas-cooled reactors (HTGRs) in the major industrial countries of the world. Existing demonstration plants and facilities are briefly described, and national programs for exploiting the unique high-temperature capabilities of the HTGR for commercial production of electricity and in process steam/heat application are discussed. (orig.)

  7. Influence of process variables on permeability and anisotropy of Biso-coated HTGR fuel particles

    International Nuclear Information System (INIS)

    Stinton, D.P.; Lackey, W.J.; Thiele, B.A.

    1977-11-01

    The effect of several important process variables on the fraction of defective particles and anisotropy of the low-temperature isotropic (LTI) coating layer was determined for Biso-coated HTGR fuel particles. Process variables considered are deposition temperature, hydrocarbon type, diluent type, and percent diluent. The effect of several other variables such as coating rate and density that depend on the process variables were also considered in this analysis. The fraction of defective particles was controlled by the dependent variables coating rate and LTI density. Coating rate was also the variable controlling the anisotropy of the LTI layer. Diluent type and diluent concentration had only a small influence on the deposition rate of the LTI layer. High-quality particles in terms of anisotropy and permeability can be produced by use of a porous plate gas distributor if the coating rate is between 3 and 5 μm/min and the coating density is between about 1.75 and 1.95 g/cm 3

  8. Waste heat gas utilization for HTGR gas turbine plant for sea water desalination

    International Nuclear Information System (INIS)

    Hunter, D.A.A.

    1981-01-01

    A thermodynamic analysis is performed for a HTGR - Gas Turbine Plant, coupled with a Rankine cycle for additional power generation and/or desalination of sea water with a multistage flash evaporator. Three basic alternatives are studied: a) Brayton cycle with inter-cooling and without regeneration, coupled with a Rankine cycle for power generation and steam for evaporator. b) Same as a) but without inter-cooling and with regeneration. c) Brayton cycle with regeneration, without inter-cooling, coupled with a Rankine cycle for sea water evaporator steam generation. The behavior of the three alternatives is established with a parametric study for the most representative variables. Economy, safety and control aspects were considered for the three different conceptions. (Author) [pt

  9. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength for the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

  10. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2008

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Tachibana, Yukio; Sun Yuliang

    2009-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2008. (author)

  11. Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2009

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Wang Hong

    2010-07-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2009. (author)

  12. Information exchange mainly on HTGR operation and maintenance technique between JAEA and INET in 2005

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Hino, Ryutaro; Yu Suyuan

    2006-06-01

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation with emphasis on HTGR operation and maintenance techniques between JAEA and INET and outlines cooperation activities during the fiscal year 2005. (author)

  13. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  14. Thermal transport properties of helium, helium--air mixtures, water, and tubing steel used in the CACHE program to compute HTGR auxiliary heat exchanger performance

    International Nuclear Information System (INIS)

    Tallackson, J.R.

    1976-02-01

    A description is presented of the thermal transport properties of the materials involved in digital computer calculations of heat transfer rates by the core auxiliary heat exchangers in large HTGR nuclear steam supply systems. These materials are pure helium, mixtures of helium with common gases having molecular weights in the range of 28 to 32, alloy steel tubing, and water. For use in programmed computations the viscosity, thermal conductivity, and specific heat are represented primarily by equations augmented by curves and tabulations. Materials supporting the development and selection of the property equations are included

  15. 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.

  16. 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

  17. Approach on a global HTGR R and D network

    International Nuclear Information System (INIS)

    Lensa, W. von

    1997-01-01

    The present situation of nuclear power in general and of the innovative nuclear reactor systems in particular requires more comprehensive, coordinated R and D efforts on a broad international level to respond to today's requirements with respect to public and economic acceptance as well as to globalization trends and global environmental problems. HTGR technology development has already reached a high degree of maturity that will be complemented by the operation of the two new test reactors in Japan and China, representing technological milestones for the demonstration of HTGR safety characteristics and Nuclear Process Heat generation capabilities. It is proposed by the IAEA 'International Working Group on Gas-Cooled Reactors' to establish a 'Global HTGR R and D Network' on basic HTGR technology for the stable, long-term advancement of the specific HTGR features and as a basis for the future market introduction of this innovative reactor system. The background and the motivation for this approach are illustrated, as well as first proposals on the main objectives, the structure and the further procedures for the implementation of such a multinational working sharing R and D network. Modern telecooperation methods are foreseen as an interactive tool for effective communication and collaboration on a global scale. (author)

  18. Study on commercial HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo

    2000-07-01

    The Japanese energy demand in 2030 will increase up to 117% in comparison with one in 2000. We have to avoid a large consumption of fossil fuel that induces a large CO 2 emission from viewpoint of global warming. Furthermore new energy resources expected to resolve global warming have difficulty to be introduced more because of their low energy density. As a result, nuclear power still has a possibility of large introduction to meet the increasing energy demand. On the other hand, in Japan, 40% of fossil fuels in the primary energy are utilized for power generation, and the remaining are utilized as a heat source. New clean energy is required to reduce the consumption of fossil fuels and hydrogen is expected as a alternative energy resource. Prediction of potential hydrogen demand in Japan is carried out and it is clarified that the demand will potentially increase up to 4% of total primary energy in 2050. In present, steam reforming method is the most economical among hydrogen generation processes and the cost of hydrogen production is about 7 to 8 yen/m 3 in Europe and the United States and about 13 yen/m 3 in Japan. JAERI has proposed for using the HTGR whose maximum core outlet temperature is at 950degC as a heat source in the steam reforming to reduced the consumption of fossil fuels and resulting CO 2 emission. Based on the survey of the production rate and the required thermal energy in conventional industry, it is clarified that a hydrogen production system by the steam reforming is the best process for the commercial HTGR nuclear heat utilization. The HTGR steam reforming system and other candidate nuclear heat utilization systems are considered from viewpoint of system layout and economy. From the results, the hydrogen production cost in the HTGR stream reforming system is expected to be about 13.5 yen/m 3 if the cost of nuclear heat of the HTGR is the same as one of the LWR. (author)

  19. Manufacture of a heat-resistant alloy with modified specifications for HTGR structural applications

    International Nuclear Information System (INIS)

    Sahira, K.; Kondo, T.; Takeiri, T.

    1984-01-01

    A method of manufacturing a nuclear grade nickel-base heat-resistant alloy in application to heliumcooled reactor primary circuit components has been developed. The Hastelloy-XR alloy, a version of Hastelloy-X, was made available by combining the basic studies of the oxidation behavior of Hastelloy-X and the improvement of manufacturing techniques. In the primary and remelting steps, the choice of appropriate processes was made by performing numerical analyses of the statistical deviation of both chemical composition and the products' mechanical properties. The feasibility of making larger electroslag remelting ingots with reasonable control of macrosegregation was examined by the calculation of a molten metal pool shape during melting. The hot workability of Hastelloy-XR was confirmed to be equivalent to that of Hastelloy-X and the importance of controlling the thermal and mechanical processes more closely was stressed in obtaining a higher level of quality assurance for the nuclear applications. The possibility of enhancing the high-temperature mechanical performance of Hastelloy-XR was suggested based on the preliminary test results with the heats manufactured with controlled boron content

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

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

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

  1. Utilization of HTGR on active carbon recycling energy system

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yukitaka, E-mail: yukitaka@nr.titech.ac.jp

    2014-05-01

    A new energy transformation concept based on carbon recycling, called as active carbon recycling energy system, ACRES, was proposed for a zero carbon dioxide emission process. The ACRES is driven availably by carbon dioxide free primary energy. High temperature gas cooled reactor (HTGR) is a candidate of the energy sources for ACRES. A smart ironmaking system with ACRES (iACRES) is one of application examples. The contribution of HTGR on iACRES was discussed thermodynamically in this study. A carbon material is re-used cyclically as energy carrier media in ACRES. Carbon monoxide (CO) had higher energy densities than hydrogen and was compatible with conventional process. Thus, CO was suitable recycling media for ACRES. Efficient regeneration of CO was a key technology for ACRES. A combined system of hydrogen production by water electrolysis and CO{sub 2} hydrogen reduction was candidate. CO{sub 2} direct electrolysis was also one of the candidates. HTGR was appropriate heat source for both water and CO{sub 2} electrolysises, and CO{sub 2} hydrogen reduction. Thermodynamic energy balances were calculated for both systems with HTGR for an ironmaking system. The direct system showed relatively advantage to the combined system in the stand point of enthalpy efficiency and simplicity of the process. One or two plants of HTGR are corresponding with ACRES system for one unit of conventional blast furnace. The proposed ACRES system with HTGR was expected to form the basis of a new energy industrial process that had low CO{sub 2} emission.

  2. HTGR safety research program

    International Nuclear Information System (INIS)

    Barsell, A.W.; Olsen, B.E.; Silady, F.A.

    1981-01-01

    An HTGR safety research program is being performed supporting and guided in priorities by the AIPA Probabilistic Risk Study. Analytical and experimental studies have been conducted in four general areas where modeling or data assumptions contribute to large uncertainties in the consequence assessments and thus, in the risk assessment for key core heat-up accident scenarios. Experimental data have been obtained on time-dependent release of fission products from the fuel particles, and plateout characteristics of condensible fission products in the primary circuit. Potential failure modes of primarily top head PCRV components as well as concrete degradation processes have been analyzed using a series of newly developed models and interlinked computer programs. Containment phenomena, including fission product deposition and potential flammability of liberated combustible gases have been studied analytically. Lastly, the behaviour of boron control material in the core and reactor subcriticality during core heatup have been examined analytically. Research in these areas has formed the basis for consequence updates in GA-A15000. Systematic derivation of future safety research priorities is also discussed. (author)

  3. Process heat applications of HTR-PM600 in Chinese petrochemical industry: Preliminary study of adaptability and economy

    International Nuclear Information System (INIS)

    Fang, Chao; Min, Qi; Yang, Yanran; Sun, Yuliang

    2017-01-01

    Highlights: •High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. •The joint of a 600 MW modular HTGR (HTR-PM600) and petrochemical industry is achievable. •The mature technology of turbine in thermal power station could be readily adopted. •The economy of this scheme is also acceptable. -- Abstract: High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. In this article, the preliminary feasibility of a 600 MW modular HTGR (HTR-PM600) working as heat source for a typical hypothetical Chinese petrochemical factory is discussed and it is found that the joint of HTR-PM600 and petrochemical industry is achievable. In detail, the heat and water balance analysis of the petrochemical factory is given. Furthermore, the direct cost of heat supplied by HTR-PM600 is calculated and corresponding economy is estimated. The results show that though there are several challenges, the application of process heat of HTGR to petrochemical industry is practical in sense of both technology and economy.

  4. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

    1985-01-01

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  5. In-line monitoring of effluents from HTGR fuel particle preparation processes using a time-of-flight mass spectrometer

    International Nuclear Information System (INIS)

    Lee, D.A.; Costanzo, D.A.; Stinton, D.P.; Carpenter, J.A.; Rainey, W.T. Jr.; Canada, D.C.; Carter, J.A.

    1976-08-01

    The carbonization, conversion, and coating processes in the manufacture of HTGR fuel particles have been studied with the use of a time-of-flight mass spectrometer. Non-condensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes which have been monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO 2 to UC 2 , buffer and low temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures

  6. HTGR fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-01-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  7. HTGR Application Economic Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  8. Effect of heat source shape on the thermal field in the pebble bed core of High Temperature Gas-cooled Reactor (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Leisheng; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

    2015-10-15

    In this study, in order to minimize the error brought by non-uniform heat flux, the spherical heaters are employed as heat source; subsequently, thermal field and heat transfer characteristics of the pebbles are investigated. The thermal field of the pebble surface in PBR is measured with heat source in different shapes. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTGR may affect the integrity of the pebbles, which has drawn the attention of many scientists to investigate the thermal field and to predict the maximum temperature locations in the pebbles using CFD method, Lee et.al has also done some experimental work on measuring the surface temperature of the pebbles as well as visualizing flow patterns of the coolant gas, and it was found that the temperature near the contacting points between pebbles was not higher than the flow stagnation points due to the higher thermal conductivity of the pebble. Certain error of temperature measurement might occur because of not very uniform heat flux in the pebbles since heater in cylindrical shape was utilized as heat source in previous experiment. More uniform heat flux and more complicated thermal profile are found in the result obtained using spherical heaters. The result shows that the temperature in contact point is higher than that in the top point, which is different from the previous results. The complex thermal phenomena observed in the lower-half side-sphere can be explained by the flow pattern near the surface.

  9. The Integration Of Process Heat Applications To High Temperature Gas Reactors

    International Nuclear Information System (INIS)

    McKellar, Michael G.

    2011-01-01

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  10. Nuclear process heat

    International Nuclear Information System (INIS)

    Barnert, H.; Hohn, H.; Schad, M.; Schwarz, D.; Singh, J.

    1993-01-01

    In a system for the application of high temperature heat from the HTR one must distinguish between the current generation and the use of process heat. In this respect it is important that the current can be generated by dual purpose power plants. The process heat is used as sensible heat, vaporisation heat and as chemical energy at the chemical conversion for the conversion of raw materials, the refinement of fossil primary energy carriers and finally circuit processes for the fission of water. These processes supply the market for heat, fuels, motor fuels and basic materials. Fifteen examples of HTR heat processes from various projects and programmes are presented in form of energy balances, however in a rather short way. (orig./DG) [de

  11. An introduction to our activities supporting HTGR developments in Japan

    International Nuclear Information System (INIS)

    An, S.; Hayashi, T.; Tsuchie, Y.

    1997-01-01

    On the view point the most important for the HTGR development promotion now in Japan is to have people know about HTGR, the Research Association of HTGR Plants(RAHP) has paid the best efforts for making an appealing report for the past two years. The outline of the report is described with an introduction of some basic experiments done on the passive decay heat removal as one of the activities carried out in a member of the association. (author)

  12. Microwave processing heats up

    Science.gov (United States)

    Microwaves are a common appliance in many households. In the United States microwave heating is the third most popular domestic heating method food foods. Microwave heating is also a commercial food processing technology that has been applied for cooking, drying, and tempering foods. It's use in ...

  13. HTGR accident and risk assessment

    International Nuclear Information System (INIS)

    Silady, F.A.; Everline, C.J.; Houghton, W.J.

    1982-01-01

    This paper is a synopsis of the high-temperature gas-cooled reactor probabilistic risk assessments (PRAs) performed by General Atomic Company. Principal topics presented include: HTGR safety assessments, peer interfaces, safety research, process gas explosions, quantitative safety goals, licensing applications of PRA, enhanced safety, investment risk assessments, and PRA design integration

  14. Design method of control system for HTGR fuel handling process with control Petri net

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

    As a complex mechanical system,the fuel handling system (FHS) of pebble-bed high temperature gas cooled reactor (HTGR) is with the features of complicated structure, numerous control devices and strict working scheduling. It is very important to precisely describe the function of FHS and effectively design its control system. A design method of control system based on control Petri net (CPN) is introduced in this paper. By associating outputs and operations with places, associating inputs and conditions with transitions, and introducing macro-places and macro-actions, the CPN realizes hierarchy design of complex control system. Based on the analysis of basic functions and working flow of FHS, its control system is described and designed by CPN. According to the firing regulation of transition,the designed CPN can be easily converted into LAD program of PLC, which can be implemented on the FHS simulating control test-bed. Application illuminates that proposed method has the advantages of clear design structure, exact description power and effective design ability of control program, which can meet the requirements of FHS control sys-tem design. (authors)

  15. Development of the krypton absorption in liquid carbon dioxide (KALC) process for HTGR off-gas reprocessing

    International Nuclear Information System (INIS)

    Glass, R.W.; Beaujean, H.W.R.; Cochran, H.D. Jr.; Haas, P.A.; Levins, D.M.; Woods, W.M.

    1975-01-01

    Reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel involves burning of the graphite-matrix elements to release the fuel for recovery purposes. The resulting off-gas is primarily CO 2 with residual amounts of N 2 , O 2 , and CO, together with fission products. Trace quantities of krypton-85 must be recovered in a concentrated form from the gas stream, but processes commonly employed for rare gas removal and concentration are not suitable for use with off-gas from graphite burning. The KALC (Krypton Absorption in Liquid CO 2 ) process employs liquid CO 2 as a volatile solvent for the krypton and is, therefore, uniquely suited to the task. Engineering development of the KALC process is currently under way at the Oak Ridge National Laboratory (ORNL) and the Oak Ridge Gaseous Diffusion Plant (ORGDP). The ORNL system is designed for close study of the individual separation operations involved in the KALC process, while the ORGDP system provides a complete pilot facility for demonstrating combined operations on a somewhat larger scale. Packed column performance and process control procedures have been of prime importance in the initial studies. Computer programs have been prepared to analyze and model operational performance of the KALC studies, and special sampling and in-line monitoring systems have been developed for use in the experimental facilities. (U.S.)

  16. HTGR Industrial Application Functional and Operational Requirements

    International Nuclear Information System (INIS)

    Demick, L.E.

    2010-01-01

    This document specifies the functional and performance requirements to be used in the development of the conceptual design of a high temperature gas-cooled reactor (HTGR) based plant supplying energy to a typical industrial facility. These requirements were developed from collaboration with industry and HTGR suppliers over the preceding three years to identify the energy needs of industrial processes for which the HTGR technology is technically and economically viable. The functional and performance requirements specified herein are an effective representation of the industrial sector energy needs and an effective basis for developing a conceptual design of the plant that will serve the broadest range of industrial applications.

  17. SC-HTGR Performance Impact for Arid Sites

    International Nuclear Information System (INIS)

    Lommers, L.; Geschwindt, J.; Southworth, F.; Shahrokhi, F.

    2014-01-01

    The SC-HTGR provides high temperature steam which can support industrial process heat applications as well as high efficiency electricity generation. The increased generating efficiency resulting from using high steam temperature provides greater plant output than lower temperature concepts, and it also reduces the fraction of waste heat which must be rejected. This capability is particularly attractive for sites with little or no water for heat rejection. This high temperature capability provides greater flexibility for these sites, and it results in a smaller performance penalty than for lower temperature systems when dry cooling must be used. The performance of the SC-HTGR for a conventional site with wet cooling is discussed first. Then the performance for arid sites is evaluated. Dry cooling performance is evaluated for both moderately arid sites and very hot sites. Offdesign performance of the dry cooling system under extreme conditions is also considered. Finally, operating strategies are explored for sites where some cooling water may be available but only in very limited quantities. Results of these assessments confirm that the higher operating temperatures of the SC-HTGR are very beneficial for arid sites, providing significant advantages for both gross and net power generation. (author)

  18. HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-79 are reported. The report covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop an MEU fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  19. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant.

  20. HTGR Generic Technology Program. Semiannual report for the period ending March 31, 1979

    International Nuclear Information System (INIS)

    1979-06-01

    This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the first half of FY-79. It covers a period when the major design direction of the National HTGR Program is in the process of changing from the HTGR-SC emphasis to an HTGR-GT emphasis in the near term. The HTGR Generic Technology Program activities have been redirected to ensure that the tasks covered are supportive of this changing emphasis in HTGR applications. The activities include the need to develop a medium enriched uranium (MEU) fuel, and the need to qualify materials and components for the higher temperatures of the gas turbine plant

  1. US HTGR Deployment Challenges and Strategies HTR 2014 Conference Proceedings

    International Nuclear Information System (INIS)

    Shahrokhi, Farshid; Lommers, Lewis; Mayer, John III; Southworth, Finis

    2014-01-01

    The NGNP Industry Alliance (NIA), LLC (www.NGNPAliance.org), is a consortium of high temperature gas-cooled reactor (HTGR) designers, utility plant owner/operators, critical plant hardware suppliers, and end-user groups. The NIA is promoting the design and commercialization of a HTGR for industrial process heat applications and electricity generation. In 2012, NIA selected the AREVA Steam Cycle HTGR (SC-HTGR) as its primary reactor design choice for its first implementation in mid -2020s. The SC-HTGR can produce 625 MWth of process steam at 550°C or 275 MWe of electricity in a co-generation configuration. The standard plant is a four-pack of 625MWth modules providing steam and electricity co-generation. The safety characteristics of the HTGR technology allows close colocation of the nuclear plant and the industrial end-user. The plant design also allows the process steam used for the industrial applications to be completely segregated and separate from primary Helium coolant and the secondary nuclear steam supply systems. The process steam at temperatures up to 550°C is provided for a variety of direct or indirect applications. End-user requirements are met for a wide range of steam flow, pressure and temperature conditions. Very high reliability (>99.99%) is maintained by the use of multi-reactor modules and conventional gas-fired back-up. Intermittent steam loads can also be efficiently met through co-generation of electricity for internal use or external distribution and sale. The NIA technology development and deployment challenges are met with strategies that provide investment and partnerships opportunities for plant design and equipment supply, and by cooperative government research, sovereign or private investment, and philanthropic opportunities. Our goal is to create intellectual property (IP) and investor value as the design matures and a license is obtained. The strategy also includes involvement of the initial customer in sharing the value created in

  2. Process heat. Triggering the processes

    Energy Technology Data Exchange (ETDEWEB)

    Augsten, Eva

    2012-07-01

    If solar process heat is to find a market, then the decision makers in industrial companies need to be aware that it actually exists. This was one of the main goals of the So-Pro project, which officially drew to a close in April 2012. (orig.)

  3. Nuclear process heat

    Energy Technology Data Exchange (ETDEWEB)

    Schulten, R [Kernforschungsanlage Juelich G.m.b.H. (F.R. Germany). Inst. fuer Reaktorentwicklung

    1976-05-01

    It is anticipated that the coupled utilization of coal and nuclear energy will achieve great importance in the future, the coal serving mainly as raw material and nuclear energy more as primary energy. Prerequisite for this development is the availability of high-temperature reactors, the state of development of which is described here. Raw materials for coupled use with nuclear process heat are petroleum, natural gas, coal, lignite, and water. Steam reformers heated by nuclear process heat, which are suitable for numerous processes, are expected to find wide application. The article describes several individual methods, all based on the transport of gas in pipelines, which could be utilized for the long distance transport of 'nuclear energy'.

  4. Universally applicable design concept of stably controlling an HTGR-hydrogen production system

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Shibata, Taiju; Nishihara, Tetsuo; Shiozawa, Shusaku

    1996-01-01

    An HTGR-hydrogen production system should be designed to have stable controllability because of a large difference in thermal dynamics between reactor and hydrogen production system and such a control design concept should be universally applicable to a variety of hydrogen production processes by the use of nuclear heat from HTGR. A transient response analysis of an HTGR-steam reforming hydrogen production system showed that a steam generator installed in a helium circuit for cooling the nuclear reactor provides stable controllability of the total system, resulting in avoiding a reactor scram. A survey of control design-related characteristics among several hydrogen production processes revealed the similarity of endothermic chemical reactions by the use of high temperature heat and that steam is required as a reactant of the endothermic reaction or for preheating a reactant. Based on these findings, a system design concept with stable controllability and universal applicability was proposed to install a steam generator as a downstream cooler of an endothermic reactor in the helium circuit of an HTGR-hydrogen production system. (author)

  5. 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

  6. GCRA perspective on the HTGR-GT plant configuration

    International Nuclear Information System (INIS)

    1979-06-01

    Design specifications for the HTGR type reactor and gas turbine combination are presented concerning the turbomachinery; generator and isophase bus duct; PCRV and internals; heat exchangers; operability; maintenance; safety and licensing; core design; and fuel design

  7. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  8. Present Status of HTGR Utilization System Development in Japan

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiaki

    2000-01-01

    Efforts are to be continuously devoted to establish and upgrade HTGR technology in the world. Japan Atomic Energy Research Institute (JAERI) has conducted the R and D of HTGRs since the 1960's in Japan, focusing on mainly the construction of High Temperature engineering Test Reactor (HTTR) which is an HTGR with a maximum helium gas temperature of 950 o C at the reactor outlet and HTGR utilization systems. The HTTR achieved first criticality on November 10, 1998 and will restart from January in 2001. In the R and D program of HTGR utilization systems, JAERI has conducted hydrogen production systems with HTGR to demonstrate the applicability of nuclear heat for extensive energy demands besides the electric power generation. JAERI has developed a hydrogen production system by steam reforming process of natural gas using nuclear heat supplied from the HTTR. Prior to the demonstration test of HTTR hydrogen production system, a 1/30-scale out-of-pile test facility is under construction for safety review and detailed design of the system. The out-of-pile test facility will be started in 2001 and will be continued about 4 years. The hydrogen permeation and corrosion tests have been carried out since 1997. Check and review for the demonstration program in the HTTR hydrogen production system will be made in 2001. Then the HTTR hydrogen production system is scheduled to be constructed from 2003 and demonstratively operated from around 2006. In parallel with the R and D of the HTTR hydrogen production system, hydrogen production method by thermochemical water splitting, so-called IS process, has been studied in JAERI. The IS process is placed as one of future candidates of the heat utilization systems of the HTTR following the steam reforming system. Continuous and stoichiometric production of hydrogen and oxygen for 48 hours was successfully achieved with a laboratory-scale apparatus mainly made of glass. Following this achievement, the study has been continued with a larger

  9. Is there a chance for commercializing the HTGR in Indonesia?

    International Nuclear Information System (INIS)

    Arbie, B.; Akhmad, Y.R.

    1997-01-01

    Indonesia is one of the developing countries in Asia-Pacific regions that actively improving or at least continuously maintain its economic growth. For this purpose, to fulfill a domestic energy demand is a vital role to achieve the goals of Indonesian development. Pertamina, the state-owned oil company, has recently called for a significant increase in domestic gas consumption in a bid to delay Indonesia becoming a net oil importer. Therefore, there is good chance for gas industry to increase their roles in generating electricity and producing automotive fuels. The latter is an interesting field of study to be correlated with the utilization of HTGR technology where the heat source could be used in the reforming process to convert natural gas into syngas as feed material in producing automotive fuels. Since the end of 1995 National Atomic Energy Agency of Indonesia (BATAN) has made an effort to increase its role in the national energy program and Batan is also able to revolve in the Giant Natuna Project or the other natural gas field projects to promote syngas production applying HTGR technology. A series of meeting with Pertamina and BPPT (the Agency for the Assessment and Application of Technology) had been performed to promote utilization of HTGR technology in the Natuna Project. In this paper governmental policy for natural gas production that may closely relate to syngas production and preliminary study for production of syngas at the Natuna Project will be discussed. It is concluded that to gain the possibility of the HTGR acceptance in the project a scenario for production and distribution should be arranged in other to achieve the break even point for automotive fuel price at about 10 US$/GJ (fuel price in 1996) in Indonesia. (author)

  10. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

    In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL) [de

  11. Heat pump augmentation of nuclear process heat

    International Nuclear Information System (INIS)

    Koutz, S.L.

    1986-01-01

    A system is described for increasing the temperature of a working fluid heated by a nuclear reactor. The system consists of: a high temperature gas cooled nuclear reactor having a core and a primary cooling loop through which a coolant is circulated so as to undergo an increase in temperature, a closed secondary loop having a working fluid therein, the cooling and secondary loops having cooperative association with an intermediate heat exchanger adapted to effect transfer of heat from the coolant to the working fluid as the working fluid passes through the intermediate heat exchanger, a heat pump connected in the secondary loop and including a turbine and a compressor through which the working fluid passes so that the working fluid undergoes an increase in temperature as it passes through the compressor, a process loop including a process chamber adapted to receive a process fluid therein, the process chamber being connected in circuit with the secondary loop so as to receive the working fluid from the compressor and transfer heat from the working fluid to the process fluid, a heat exchanger for heating the working fluid connected to the process loop for receiving heat therefrom and for transferring heat to the secondary loop prior to the working fluid passing through the compressor, the secondary loop being operative to pass the working fluid from the process chamber to the turbine so as to effect driving relation thereof, a steam generator operatively associated with the secondary loop so as to receive the working fluid from the turbine, and a steam loop having a feedwater supply and connected in circuit with the steam generator so that feedwater passing through the steam loop is heated by the steam generator, the steam loop being connected in circuit with the process chamber and adapted to pass steam to the process chamber with the process fluid

  12. HTGR market assessment: interim report

    International Nuclear Information System (INIS)

    1979-09-01

    The purpose of this Assessment is to establish the utility perspective on the market potential of the HTGR. The majority of issues and conclusions in this report are applicable to both the HTGR-Gas Turbine (GT) and the HTGR-Steam Cycle (SC). This phase of the HTGR Market Assessment used the HTGR-GT as the reference design as it is the present focus of the US HTGR Program. A brief system description of the HTGR-GT is included in Appendix A. This initial report provides the proposed structure for conducting the HTGR Market Assessment plus preliminary analyses to establish the magnitude and nature of key factors that affect the HTGR market. The HTGR market factors and their relationship to the present HTGR Program are discussed. This report discusses two of these factors in depth: economics and water availability. The water availability situation in the US and its impact on the potential HTGR market are described. The approach for applying the HTGR within a framework of utility systems analyses is presented

  13. HTGR Application Economic Model Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  14. Reprocessing yields and material throughput: HTGR recycle demonstration facility

    International Nuclear Information System (INIS)

    Holder, N.; Abraham, L.

    1977-08-01

    Recovery and reuse of residual U-235 and bred U-233 from the HTGR thorium-uranium fuel cycle will contribute significantly to HTGR fuel cycle economics and to uranium resource conservation. The Thorium Utilization National Program Plan for HTGR Fuel Recycle Development includes the demonstration, on a production scale, of reprocessing and refabrication processes in an HTGR Recycle Demonstration Facility (HRDF). This report addresses process yields and material throughput that may be typically expected in the reprocessing of highly enriched uranium fuels in the HRDF. Material flows will serve as guidance in conceptual design of the reprocessing portion of the HRDF. In addition, uranium loss projections, particle breakage limits, and decontamination factor requirements are identified to serve as guidance to the HTGR fuel reprocessing development program

  15. HTGR-INTEGRATED COAL TO LIQUIDS PRODUCTION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Anastasia M Gandrik; Rick A Wood

    2010-10-01

    As part of the DOE’s Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to “shift” the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700°C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: • 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal

  16. HTGR-Integrated Coal To Liquids Production Analysis

    International Nuclear Information System (INIS)

    Gandrik, Anastasia M.; Wood, Rick A.

    2010-01-01

    As part of the DOE's Idaho National Laboratory (INL) nuclear energy development mission, the INL is leading a program to develop and design a high temperature gas-cooled reactor (HTGR), which has been selected as the base design for the Next Generation Nuclear Plant. Because an HTGR operates at a higher temperature, it can provide higher temperature process heat, more closely matched to chemical process temperatures, than a conventional light water reactor. Integrating HTGRs into conventional industrial processes would increase U.S. energy security and potentially reduce greenhouse gas emissions (GHG), particularly CO2. This paper focuses on the integration of HTGRs into a coal to liquids (CTL) process, for the production of synthetic diesel fuel, naphtha, and liquefied petroleum gas (LPG). The plant models for the CTL processes were developed using Aspen Plus. The models were constructed with plant production capacity set at 50,000 barrels per day of liquid products. Analysis of the conventional CTL case indicated a potential need for hydrogen supplementation from high temperature steam electrolysis (HTSE), with heat and power supplied by the HTGR. By supplementing the process with an external hydrogen source, the need to 'shift' the syngas using conventional water-gas shift reactors was eliminated. HTGR electrical power generation efficiency was set at 40%, a reactor size of 600 MWth was specified, and it was assumed that heat in the form of hot helium could be delivered at a maximum temperature of 700 C to the processes. Results from the Aspen Plus model were used to perform a preliminary economic analysis and a life cycle emissions assessment. The following conclusions were drawn when evaluating the nuclear assisted CTL process against the conventional process: (1) 11 HTGRs (600 MWth each) are required to support production of a 50,000 barrel per day CTL facility. When compared to conventional CTL production, nuclear integration decreases coal consumption by 66

  17. Solar Process Heat Basics | NREL

    Science.gov (United States)

    Process Heat Basics Solar Process Heat Basics Commercial and industrial buildings may use the same solar technologies-photovoltaics, passive heating, daylighting, and water heating-that are used for residential buildings. These nonresidential buildings can also use solar energy technologies that would be

  18. FY 1981 HTGR program summary-level program outline (revision 1/30/81)

    International Nuclear Information System (INIS)

    1981-01-01

    The objective of the DOE HTGR Program is the development of technology for the most important HTGR applications. Through this support, DOE seeks to encourage private sector initiatives which will lead to the development of commercially attractive HTGR applications that concurrently support national energy goals. Currently perceived as important to national energy goals are applications that primarily address the process heat market with a view toward reduction of national requirements for oil, natural gas and coal. A high priority during FY 1981, therefore, will be to further identify and define the details of the Technology Program so as to assure that it is both necessary and sufficient to provide the required support. In the establishment of a supportive Technology Program, key elements which will be addressed are as follows: studies will be conducted to further identify and characterize important unique HTGR applications and to evaluate their potential in the context of market opportunities, utility/user interest, and national objectives to develop new energy supply options; based upon the configurations and operating characteristics projected for selected applications, Technology Program requirements must be identified to support development, verification, and ultimately licensing of components and systems comprising the facilities of interest; and in the context of limited resources, sufficient analysis and evaluation must be accomplished so as to prioritize technology elements in accordance with appropriately developed criteria

  19. Cesium transport data for HTGR systems

    International Nuclear Information System (INIS)

    Myers, B.F.; Bell, W.E.

    1979-09-01

    Cesium transport data on the release of cesium from HTGR fuel elements are reviewed and discussed. The data available through 1976 are treated. Equations, parameters, and associated variances describing the data are presented. The equations and parameters are in forms suitable for use in computer codes used to calculate the release of metallic fission products from HTGR fuel elements into the primary circuit. The data cover the following processes: (1) diffusion of cesium in fuel kernels and pyrocarbon, (2) sorption of cesium on fuel rod matrix material and on graphite, and (3) migration of cesium in graphite. The data are being confirmed and extended through work in progress

  20. Factors affecting defective fraction of biso-coated HTGR fuel particles during in-block carbonization

    International Nuclear Information System (INIS)

    Caputo, A.J.; Johnson, D.R.; Bayne, C.K.

    1977-01-01

    The performance of Biso-coated thoria fuel particles during the in-block processing step of HTGR fuel element refabrication was evaluated. The effect of various process variables (heating rate, particle crushing strength, horizontal and/or vertical position in the fuel element blocks, and fuel hole permeability) on pitch coke yield, defective fraction of fuel particles, matrix structure, and matrix porosity was evaluated. Of the variables tested, only heating rate had a significant effect on pitch coke yield while both heating rate and particle crushing strength had a significant effect on defective fraction of fuel particles

  1. HTGR technology development in Japan advances so much. Leading world technology to global standards

    International Nuclear Information System (INIS)

    Ogawa, Masuro; Hino, Ryutaro; Kunitomi, Kazuhiko; Onuki, Kaoru; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Sawa, Kazuhiro

    2007-01-01

    The JAEA has conducted research and development of HTGR for hydrogen production since 1969 and attained the operation of 950degC at reactor coolant outlet of the HTTR in 2004. This article describes present status and future plan of R and D in the area of HTGR technology and high temperature heat utilization and also introduces the design of the commercial HTGR cogeneration system based on R and D results leading to world standards. (T. Tanaka)

  2. National HTGR safety program

    International Nuclear Information System (INIS)

    Davis, D.E.; Kelley, A.P. Jr.

    1982-01-01

    This paper presents an overview of the National HTGR Program in the US with emphasis on the safety and licensing strategy being pursued. This strategy centers upon the development of an integrated approach to organizing and classifying the functions needed to produce safe and economical nuclear power production. At the highest level, four plant goals are defined - Normal Operation, Core and Plant Protection, Containment Integrity and Emergency Preparedness. The HTGR features which support the attainment of each goal are described and finally a brief summary is provided of the current status of the principal safety development program supporting the validation of the four plant goals

  3. Designing heat exchangers for process heat reactors

    International Nuclear Information System (INIS)

    Quade, R.N.

    1980-01-01

    A brief account is given of the IAEA specialist meeting on process heat applications technology held in Julich, November 1979. The main emphasis was on high temperature heat exchange. Papers were presented covering design requirements, design construction and prefabrication testing, and selected problems. Primary discussion centered around mechanical design, materials requirements, and structural analysis methods and limits. It appears that high temperature heat exchanges design to nuclear standards, is under extensive development but will require a lengthy concerted effort before becoming a commercial reality. (author)

  4. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    Science.gov (United States)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

  5. Assessment of the licensing aspects of HTGR in Yugoslavia

    International Nuclear Information System (INIS)

    Varazdinec, Z.

    1990-01-01

    This paper deals not only with the licensing procedure in Yugoslavia, but also reflects the Utility/Owner approach to the assessment of the licensability of the HTGR during the site selection process and especially during bid evaluation process. Besides the description of the existing procedure which was implemented on licensing of LWR program, the assessment of some licensing aspects of HTGR has been presented to describe possible implementation on licensing procedure. (author)

  6. Assessment of the licensing aspects of HTGR in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Varazdinec, Z [Institut za Elektroprivredu-Zagreb, Zagreb (Yugoslavia)

    1990-07-01

    This paper deals not only with the licensing procedure in Yugoslavia, but also reflects the Utility/Owner approach to the assessment of the licensability of the HTGR during the site selection process and especially during bid evaluation process. Besides the description of the existing procedure which was implemented on licensing of LWR program, the assessment of some licensing aspects of HTGR has been presented to describe possible implementation on licensing procedure. (author)

  7. HTGR gas turbine program. Semiannual progress report, April 1-September 30, 1978

    International Nuclear Information System (INIS)

    1979-12-01

    This report describes work performed under the gas turbine HTGR (HTGR-GT) program, Department of Energy Contract DE-AT03-76-SF70046, during the period April 1, 1978 through September 30, 1978. The work reported covers the demonstration and commercial plant concept studies including plant layout, heat exchanger studies, turbomachine studies, systems analysis, and reactor core engineering

  8. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

  9. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

  10. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  11. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

  12. Small demonstration HTGR concept

    International Nuclear Information System (INIS)

    Kiryushin, A.I.

    1989-01-01

    Currently the USSR is investigating two high-temperature gas-cooled reactors. The first plant is the VGM, a modular type HTGR with power rating of 180-250 MWth. The second plant is the VG-400 with 1000 MWth and a prestressed concrete reactor vessel. The paper contains the description of the VGM design and its main components. (author). 1 fig., 1 tab

  13. Steam generator design considerations for modular HTGR plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; DeFur, D.D.

    1986-01-01

    Studies are in progress to develop a standard High Temperature Gas-Cooled Reactor (HTGR) plant design that is amenable to serial production and is licensable. Based on the results of trade studies performed in the DOE-funded HTGR program, activities are being focused to emphasize a modular concept based on a 350 MW(t) annular reactor core with prismatic fuel elements. Utilization of a multiplicity of the standard module affords flexibility in power rating for utility electricity generation. The selected modular HTGR concept has the reactor core and heat transport systems housed in separate steel vessels. This paper highlights the steam generator design considerations for the reference plant, and includes a discussion of the major features of the heat exchanger concept and the technology base existing in the U.S

  14. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1987-01-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

  15. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  16. An overview of possible High Temperature Gas-cooled Reactors - Gas Turbine (HTGR-GT) systems for the production of electricity and heat. Includes a technical assessment of the suitability for a small Dutch cogeneration plant; Een overzicht van mogelijke HTGR-GT systemen voor produktie van elektriciteit en warmte. Met technische beoordeling van geschiktheid voor een kleine Nederlandse W/K centrale

    Energy Technology Data Exchange (ETDEWEB)

    Kikstra, J.F

    1997-06-01

    There is a large number of different configurations for the combination of a closed cycle gas turbine (CCGT) system and a high-temperature gas-cooled reactor (HTGR). Based on the results of a literature survey an overview of such configurations is presented and a comparison is made for their appropriateness for a small cogeneration system (<60 MWt) to be used in the Netherlands. However, most cycles can only be applied for large-scale energy production or supply heat on a too low temperature level. The direct, recuperated cycle is the only suitable cycle, while that cycle is a simple system and shows an acceptable electric and total efficiency. Calculations were carried out for the co-production of hot water (75-125C and 40-70C) and for steam (10 bar, 220C). By means of a static model and an optimizer the feasible efficiencies for different heat demand are determined. The maximum electric efficiency is 42% for the co-production of hot water and 38% for the co-production of steam. 28 refs.

  17. Improvements of reforming performance of a nuclear heated steam reforming process

    International Nuclear Information System (INIS)

    Hada, Kazuhiko

    1996-10-01

    Performance of an energy production process by utilizing high temperature nuclear process heat was not competitive to that by utilizing non-nuclear process heat, especially fossil-fired process heat due to its less favorable chemical reaction conditions. Less favorable conditions are because a temperature of the nuclear generated heat is around 950degC and the heat transferring fluid is the helium gas pressurized at around 4 MPa. Improvements of reforming performance of nuclear heated steam reforming process were proposed in the present report. The steam reforming process, one of hydrogen production processes, has the possibility to be industrialized as a nuclear heated process as early as expected, and technical solutions to resolve issues for coupling an HTGR with the steam reforming system are applicable to other nuclear-heated hydrogen production systems. The improvements are as follows: As for the steam reformer, (1) increase in heat input to process gas by applying a bayonet type of reformer tubes and so on, (2) increase in reforming temperature by enhancing heat transfer rate by the use of combined promoters of orifice baffles, cylindrical thermal radiation pipes and other proposal, and (3) increase in conversion rate of methane to hydrogen by optimizing chemical compositions of feed process gas. Regarding system arrangement, a steam generator and superheater are set in the helium loop as downstream coolers of the steam reformer, so as to effectively utilize the residual nuclear heat for generating feed steam. The improvements are estimated to achieve the hydrogen production rate of approximately 3800 STP-m 3 /h for the heat source of 10 MW and therefore will provide the potential competitiveness to a fossil-fired steam reforming process. Those improvements also provide the compactness of reformer tubes, giving the applicability of seamless tubes. (J.P.N.)

  18. The Thermos process heat reactor

    International Nuclear Information System (INIS)

    Lerouge, Bernard

    1979-01-01

    The THERMOS process heat reactor was born from the following idea: the hot water energy vector is widely used for heating purposes in cities, so why not save on traditional fossil fuels by simply substituting a nuclear boiler of comparable power for the classical boiler installed in the same place. The French Atomic Energy Commission has techniques for heating in the big French cities which provide better guarantees for national independence and for the environment. This THERMOS technique would result in a saving of 40,000 to 80,000 tons of oil per year [fr

  19. Time-dependent high-temperature low-cycle fatigue behavior of nickel-base heat-resistant alloys for HTGR

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Kondo, Tatsuo

    1988-06-01

    A series of strain controlled low-cycle fatigue tests at 900 deg C in the simulated HTGR helium environment were conducted on Hastelloy X and its modified version, Hastelloy XR in order to examine time-dependent high-temperature low-cycle fatigue behavior. In the tests with the symmetric triangular strain waveform, decreasing the strain rate led to notable reductions in the fatigue life. In the tests with the trapezoidal strain waveform with different holding types, the fatigue life was found to be reduced most effectively in tensile hold-time experiments. Based on the observations of the crack morphology the strain holding in the compressive side was suggested to play the role of suppressing the initiation and the growth of internal cracks or cavities, and to cause crack branching. When the frequency modified fatigue life method and/or the prediction of life by use of the ductility were applied, both the data obtained with the symmetric triangular strain waveform and those with the tensile hold-time experiments lay on the straight line plots. The data, however, obtained with the compressive and/or both hold-time experiments could not be handled satisfactorily by those methods. When the cumulative damage rule was applied, it was found that the reliability of HTGR components was ensured by limiting the creep-fatigue damage fraction within the value of 1. (author)

  20. The HTTR project as the world leader of HTGR research and development

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

    2005-01-01

    As a next generation type nuclear system which will expand nuclear energy use area with high temperature nuclear heat utilization and improve economic competitiveness greatly, High Temperature Gas-cooled Reactor (HTGR) has become the R and D item of prime importance at home as well as abroad to establish hydrogen society to cope with global environmental problems. JAERI has conducted R and D on HTGR as the world leader such as to achieve a reactor outlet coolant temperature of 950 degC in the HTTR (High Temperature Engineering Test Reactor) in April 2004 as the world's first and also to succeed in continuous hydrogen production with a bench-scale apparatus of closed cycle iodine-sulfur (IS) process for six and half hours in August 2003 as the world's first. Overview and present status of HTTR program were presented in details with background and main R and D results as well as international trend of HTGR development and future program on pilot tests facilities for hydrogen production demonstration in Japan. (T. Tanaka)

  1. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  2. An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

    2008-03-01

    To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

  3. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Baxter, B.; Benedict, G.E.; Zimmerman, R.D.

    1976-01-01

    Development studies to date indicate that the HTGR fuel blocks can be effectively crushed with two stages of eccentric jaw crushing, followed by a double-roll crusher, a screener and an eccentrically mounted single-roll crusher for oversize particles. Burner development results indicate successful long-term operation of both the primary and secondary fluidized-bed combustion systems can be performed with the equipment developed in this program. Aqueous separation development activities have centered on adapting known Acid-Thorex processing technology to the HTGR reprocessing task. Significant progress has been made on dissolution of burner ash, solvent extraction feed preparation, slurry transfer, solids drying and solvent extraction equipment and flowsheet requirements

  4. Exergy analysis of HTGR-GT

    International Nuclear Information System (INIS)

    Cao Jianhua; Wang Jie; Yang Xiaoyong; Yu Suyuan

    2005-01-01

    The High Temperature Gas-cooled Reactor (HTGR) coupled with gas turbine for high efficiency in electricity production is supposed to be one of the candidates for the future nuclear power plants. The HTGR gas turbine cycle is theoretically based on the Brayton cycle with recuperated, intercooled and precooled sub-processes. In this paper, an exergy analysis of the Brayton Cycle on HTGR is presented. The analyses were done for four typical reactor outlet temperatures and the exergy loss distribution and exergy loss ratio of each sub-process was quantified. The results show that more than a half of the exergy loss takes place in the reactor, while the low pressure compressor (LPC), the high pressure compressor (HPC) and the intercooler denoted by compress system together, play a much small role in the contribution of exergy losses. With the rise of the reactor outlet temperature, both the exergy loss and exergy loss ratio of the reactor can be greatly cut down, so is the total exergy loss of the cycle; while the exergy loss ratios of the recuperator and precooler have a small rise. The total exergy efficiency of the cycle is quite high (50% more or less). (authors)

  5. The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

    Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

  6. Status of international HTGR development

    International Nuclear Information System (INIS)

    Homan, F.J.; Simon, W.A.

    1988-01-01

    Programs for the development of high-temperature gas-cooled reactor (HTGR) technology over the past 30 years in eight countries are briefly described. These programs have included both government sector and industrial sector participation. The programs have produced four electricity-producing prototype/demonstration reactors, two in the United States, and two in the Federal Republic of Germany. Key design parameters for these ractors are compared with the design parameters planned for follow-on commercial-scale HTGRs. The development of HTGR technology has been enhanced by numerous cooperative agreements over the years, involving both government-sponsored national laboratories and industrial participants. Current bilateral cooperative agreements are described. A relatively new component in the HTGR international cooperation is that of multinational industrial alliances focused on supplying commercial-scale HTGR power plants. Current industrial cooperative agreements are briefly discussed

  7. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

    An overview of HTGR fuel recycle is presented, with emphasis placed on reprocessing and fuel kernel refabrication. Overall recycle operations include (1) shipment and storage, (2) reprocessing, (3) refabrication, (4) waste handling, and (5) accountability and safeguards

  8. Selection of JAERI'S HTGR-GT concept

    International Nuclear Information System (INIS)

    Muto, Y.; Ishiyama, S.; Shiozawa, S.

    2001-01-01

    In JAERI, a feasibility study of HTGR-GT has been conducted as an assigned work from STA in Japan since January 1996. So far, the conceptual or preliminary designs of 600, 400 and 300 MW(t) power plants have been completed. The block type core and pebble-bed core have been selected in 600 MW(t) and 400/300 MW(t), respectively. The gas-turbine system adopts a horizontal single shaft rotor and then the power conversion vessel is separated into a turbine vessel and a heat exchanger vessel. In this paper, the issues related to the selection of these concepts are technically discussed. (author)

  9. HTGR Cost Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooler Reactor (HTGR) Cost Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Cost Model calculates an estimate of the capital costs, annual operating and maintenance costs, and decommissioning costs for a high-temperature gas-cooled reactor. The user can generate these costs for multiple reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for a single or four-pack configuration; and for a reactor size of 350 or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Cost Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Cost Model. This model was design for users who are familiar with the HTGR design and Excel. Modification of the HTGR Cost Model should only be performed by users familiar with Excel and Visual Basic.

  10. Study on the inspection item and inspection method of HTGR fuel

    International Nuclear Information System (INIS)

    Na, Sang Ho; Kim, Y. K.; Jeong, K. C.; Oh, S. C.; Cho, M. S.; Kim, Y. M.; Lee, Y. W.

    2006-01-01

    The type of HTGR(High Temperature Gas-cooled Reactor) fuel is different according to the reactor type. Generally the HTGR fuel has two types. One is a block type, which is manufactured in Japan or America. And the other is a pebble type, which is manufactured in China. Regardless of the fuel type, the fuel manufacturing process started from the coated particle, which is consisted of fuel kernel and the 4 coating layers. Korea has a plan to fabricate a HTGR fuel in near future. The appropriate quality inspection standards are requested to produce a sound and reliable coated particle for HTGR fuel. Therefore, the inspection items and the inspection methods of HTGR fuel between Japan and China, which countries have the manufacturing process, are investigated to establish a proper inspection standards of our product characteristics

  11. High-temperature gas reactor (HTGR) market assessment, synthetic fuels analysis

    International Nuclear Information System (INIS)

    1980-08-01

    This study is an update of assessments made in TRW's October 1979 assessment of overall high-temperature gas-cooled reactor (HTGR) markets in the future synfuels industry (1985 to 2020). Three additional synfuels processes were assessed. Revised synfuel production forecasts were used. General environmental impacts were assessed. Additional market barriers, such as labor and materials, were researched. Market share estimates were used to consider the percent of markets applicable to the reference HTGR size plant. Eleven HTGR plants under nominal conditions and two under pessimistic assumptions are estimated for selection by 2020. No new HTGR markets were identified in the three additional synfuels processes studied. This reduction in TRW's earlier estimate is a result of later availability of HTGR's (commercial operation in 2008) and delayed build up in the total synfuels estimated markets. Also, a latest date for HTGR capture of a synfuels market could not be established because total markets continue to grow through 2020. If the nominal HTGR synfuels market is realized, just under one million tons of sulfur dioxide effluents and just over one million tons of nitrous oxide effluents will be avoided by 2020. Major barriers to a large synfuels industry discussed in this study include labor, materials, financing, siting, and licensing. Use of the HTGR intensifies these barriers

  12. HTGR safety research concerns at NRC

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1982-01-01

    A general discussion of HTGR technical and safety-related problems is given. The broad areas of current research programs specific to the Fort St. Vrain reactor and applicable to HTGR technology are summarized

  13. Nuclear energy and process heating

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-10-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating and several industrial applications. Although only About 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven, a

  14. Nuclear energy and process heating

    International Nuclear Information System (INIS)

    Kozier, K.S.

    1999-10-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating and several industrial applications. Although only About 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven, a determined

  15. Status of CHAP: composite HTGR analysis program

    International Nuclear Information System (INIS)

    Secker, P.A.; Gilbert, J.S.

    1975-12-01

    Development of an HTGR accident simulation program is in progress for the prediction of the overall HTGR plant transient response to various initiating events. The status of the digital computer program named CHAP (Composite HTGR Analysis Program) as of June 30, 1975, is given. The philosophy, structure, and capabilities of the CHAP code are discussed. Mathematical descriptions are given for those HTGR components that have been modeled. Component model validation and evaluation using auxiliary analysis codes are also discussed

  16. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  17. USNRC HTGR safety research program overview

    International Nuclear Information System (INIS)

    Foulds, R.B.

    1982-01-01

    An overview is given of current activities and planned research efforts of the US Nuclear Regulatory Commission (NRC) HTGR Safety Program. On-going research at Brookhaven National Laboratory, Oak Ridge National Laboratory, Los Alamos National Laboratory, and Pacific Northwest Laboratory are outlined. Tables include: HTGR Safety Issues, Program Tasks, HTGR Computer Code Library, and Milestones for Long Range Research Plan

  18. HTGR analytical methods and design verification

    International Nuclear Information System (INIS)

    Neylan, A.J.; Northup, T.E.

    1982-05-01

    Analytical methods for the high-temperature gas-cooled reactor (HTGR) include development, update, verification, documentation, and maintenance of all computer codes for HTGR design and analysis. This paper presents selected nuclear, structural mechanics, seismic, and systems analytical methods related to the HTGR core. This paper also reviews design verification tests in the reactor core, reactor internals, steam generator, and thermal barrier

  19. HTGR [High Temperature Gas-Cooled Reactor] ingress analysis using MINET

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs

  20. Personnel radiation exposure in HTGR plants

    International Nuclear Information System (INIS)

    Su, S.; Engholm, B.A.

    1981-01-01

    Occupational radiation exposures in high-temperature gas-cooled reactor (HTGR) plants were assessed. The expected rate of dose accumulations for a large HTGR steam cycle unit is 0.07 man-rem/MW(e)y, while the design basis is 0.17 man-rem/MW(e)y. The comparable figure for actual light water reactor experience is 1.3 man-rem/MW(e)y. The favorable HTGR occupational exposure is supported by results from the Peach Bottom Unit No. 1 HTGR and Fort St. Vrain HTGR plants and by operating experience at British gas-cooled reactor stations

  1. Development, design, and preliminary operation of a resin-feed processing facility for resin-based HTGR fuels

    International Nuclear Information System (INIS)

    Haas, P.A.; Drago, J.P.; Million, D.L.; Spence, R.D.

    1978-01-01

    Fuel kernels for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared by loading carboxylic acid cation exchange resins with uranium and carbonizing at controlled conditions. Resin-feed processing was developed and a facility was designed, installed, and operated to control the kernel size, shape, and composition by processing the resin before adding uranium. The starting materials are commercial cation exchange resins in the sodium form. The size separations are made by vibratory screening of resin slurries in water. After drying in a fluidized bed, the nonspherical particles are separated from spherical particles on vibratory plates of special design. The sized, shape-separated spheres are then rewetted and converted to the hydrogen form. The processing capacity of the equipment tested is equivalent to about 1 kg of uranium per hour and could meet commercial recycle plant requirements without scale-up of the principal process components

  2. Summary of foreign HTGR programs

    International Nuclear Information System (INIS)

    1980-06-01

    This report contains pertinent information on the status, objectives, budgets, major projects and facilities, as well as user, industrial and governmental organizations involved in major foreign gas-cooled thermal reactor programs. This is the second issue of this document (the first was issued in March 1979). The format has been revised to consolidate material according to country. These sections are followed by the foreign HTGR program index which serves as a quick reference to some of the many acronyms associated with the foreign HTGR programs

  3. HTGR R and D programs

    International Nuclear Information System (INIS)

    Neylan, A.J.; Brisbois, J.

    1979-01-01

    A significant R and D program (including in certain cases full-scale prototype tests) formed the basis for the design and key elements in the foregoing projects and is continuing to provide a basis for generic design development. HTGR R and D programs are both privately and government sponsored. This paper provides an overview of the background, current status and outstanding design issues/problems remaining in the area of NSS Plant, Materials and Fuel. The specific objectives and scope of all recently completed, ongoing and planned major HTGR R and D programs are presented

  4. Present status of HTGR research and development, 1995

    International Nuclear Information System (INIS)

    1996-02-01

    Based on the Long-term Program for Development and Utilization of Nuclear Energy which was revised in 1987, the Japan Atomic Energy Research Institute (JAERI) has carried out the Research and Development (R and D) on the High Temperature Gas-cooled Reactors (HTGRs) in Japan. The JAERI obtained the installation permit of the High Temperature Engineering Test Reactor (HTTR) from the Government in November 1990 and started the construction of the HTTR facility in the Oarai Research Establishment in March 1991. The HTTR is a test reactor with thermal output of 30MW and outlet coolant temperature of 850degC at the rated operation and 950degC at the high temperature test operation, using the pin-in-block type fuel, and has capability to demonstrate nuclear process heat utilization. The reactor pressure vessel and intermediate heat exchanger were installed in the reactor containment vessel in 1994, and reactor internals were also installed in the reactor pressure vessel in 1995. The first criticality will be attained in December 1997. This report describes the design outline and construction progress of the HTTR, R and D of fuel, materials and components for the HTGR and high temperature nuclear heat application, and innovative and basic researches for high temperature technologies at the HTTR. (J.P.N.)

  5. Experiments Demonstrate Geothermal Heating Process

    Science.gov (United States)

    Roman, Harry T.

    2012-01-01

    When engineers design heat-pump-based geothermal heating systems for homes and other buildings, they can use coil loops buried around the perimeter of the structure to gather low-grade heat from the earth. As an alternative approach, they can drill well casings and store the summer's heat deep in the earth, then bring it back in the winter to warm…

  6. HTGR Economic / Business Analysis and Trade Studies Market Analysis for HTGR Technologies and Applications

    Energy Technology Data Exchange (ETDEWEB)

    Richards, Matt [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States); Hamilton, Chris [Ultra Safe Nuclear Corporation, Los Alamos, NM (United States)

    2013-11-01

    This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal to liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.

  7. Characterization of industrial process waste heat and input heat streams

    Energy Technology Data Exchange (ETDEWEB)

    Wilfert, G.L.; Huber, H.B.; Dodge, R.E.; Garrett-Price, B.A.; Fassbender, L.L.; Griffin, E.A.; Brown, D.R.; Moore, N.L.

    1984-05-01

    The nature and extent of industrial waste heat associated with the manufacturing sector of the US economy are identified. Industry energy information is reviewed and the energy content in waste heat streams emanating from 108 energy-intensive industrial processes is estimated. Generic types of process equipment are identified and the energy content in gaseous, liquid, and steam waste streams emanating from this equipment is evaluated. Matchups between the energy content of waste heat streams and candidate uses are identified. The resultant matrix identifies 256 source/sink (waste heat/candidate input heat) temperature combinations. (MHR)

  8. Identification of domestic needs of modular HTR for electric and heat process industry in Indonesia

    International Nuclear Information System (INIS)

    Rusli, A.; Arbie, B.

    2000-01-01

    Identification of potential applications of the modular high temperature gas-cooled reactor (HTGR) in Indonesia has been carried out his was done by surveying and analysing the electric and industrial heat process which included captive power for household and industrial complexes in 13 regional operation areas covered by PT PLN National Electric Company. The area includes from the western Aceh to the eastern Irian Jaya, cities which are 6000 miles apart. Surveying was conducted for several parameters that electric and process heat demand in each region including captive power, geological characteristics of the region, distance of each region from conventional energy resources and the existence of petrochemical industries or other industries which use high temperatures (>500 deg C). In order to obtain a scale of priority in each region, credit points (1-3) and sensitivity factors (0-1) were applied to obtain the total significant value. The regions included are: the east cost of Sumatra, the north coast of Java, the west, south and east coat of Kalimantan as well as the south coast of Sulawesi, beyond which there are thousands of small islands that are safe from a geological and a tectonic point of view. In the preliminary survey, the analysation showed that Regions III, IV, XII and XIII have a high potential priority, followed by Regions I, II, VI, VIII and IX. The potential of domestic participation in the first and second unit was also investigated in relation to the possibility of implementing the HTGR project in Indonesia in the future, and it amounted to about 26% and 31% of the total cost for the first and second units, respectively. A summary of results is shown in Table 1. (authors)

  9. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  10. Analysis of some accident conditions in confirmation of the HTGR safety

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Grishanin, E.I.; Kukharkin, N.E.; Mikhailov, P.V.; Pinchuk, V.V.; Ponomarev-Stepnoy, N.N.; Fedin, G.I.; Shilov, V.N.; Yanushevich, I.V.

    1981-01-01

    This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved

  11. Analysis of some accident conditions in confirmation of the HTGR safety

    Energy Technology Data Exchange (ETDEWEB)

    Grebennik, V. N.; Grishanin, E. I.; Kukharkin, N. E.; Mikhailov, P. V.; Pinchuk, V. V.; Ponomarev-Stepnoy, N. N.; Fedin, G. I.; Shilov, V. N.; Yanushevich, I. V. [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

    1981-01-15

    This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved.

  12. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  13. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented

  14. Heat Transfer in a Thermoacoustic Process

    Science.gov (United States)

    Beke, Tamas

    2012-01-01

    Thermoacoustic instability is defined as the excitation of acoustic modes in chambers with heat sources due to the coupling between acoustic perturbations and unsteady heat addition. The major objective of this paper is to achieve accurate theoretical results in a thermoacoustic heat transfer process. We carry out a detailed heat transfer analysis…

  15. Automated-process gas-chromatograph system for use in accelerated corrosion testing of HTGR core-support posts

    International Nuclear Information System (INIS)

    Harper, R.E.; Herndon, P.G.

    1982-01-01

    An automated-process gas chromatograph is the heart of a gaseous-impurities-analysis system developed for the Oak Ridge National Laboratory Core Support Performance Test, at which graphite core-support posts for high-temperature gas-cooled fission reactors are being subjected to accelerated corrosion tests under tightly controlled conditions of atmosphere and temperature. Realistic estimation of in-core corrosion rates is critically dependent upon the accurate measurement of low concentrations of CO, CO 2 , CH 4 , H 2 , and O 2 in the predominantly helium atmosphere. In addition, the capital and labor investment associated with each test puts a premium upon the reliability of the analytical system, as excessive downtime or failure to obtain accurate data would result in unacceptable costs and schedule delays. After an extensive survey of available measurement techniques, gas chromatography was chosen for reasons of accuracy, flexibility, good-performance record, and cost

  16. Modeling of Heating During Food Processing

    Science.gov (United States)

    Zheleva, Ivanka; Kamburova, Veselka

    Heat transfer processes are important for almost all aspects of food preparation and play a key role in determining food safety. Whether it is cooking, baking, boiling, frying, grilling, blanching, drying, sterilizing, or freezing, heat transfer is part of the processing of almost every food. Heat transfer is a dynamic process in which thermal energy is transferred from one body with higher temperature to another body with lower temperature. Temperature difference between the source of heat and the receiver of heat is the driving force in heat transfer.

  17. Phase change heat transfer device for process heat applications

    International Nuclear Information System (INIS)

    Sabharwall, Piyush; Patterson, Mike; Utgikar, Vivek; Gunnerson, Fred

    2010-01-01

    The next generation nuclear plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼1300 K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+ m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a thermal device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via 'pumping a fluid', a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization/condensing process. The condensate is further returned to the hot source by gravity, i.e., without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.

  18. Study on the Efficient Disintegration of HTGR Fuel Elements by Electrochemical Method

    International Nuclear Information System (INIS)

    Piao Nan; Chen Ji; Xiao Cuiping; We Mingfen; Che Jing

    2014-01-01

    The spent fuel elements in High- temperature gas-cooled reactor (HTGR) have a special structure, so the head-end process of the spent fuel reprocessing is different from the process of water reactor spent fuel. The first step of head-end process of the HTGR spent fuel reprocessing process is disintegration of the graphite matrix and separation of the coated fuel particles. Electrochemical method with nitrate solution as an electrolyte for fuel element disintegration has been conducted by the Institute of Nuclear and New Energy Technology in Tsinghua University. This method allows a total disintegration of graphite matrix, while still preserving the integrity of TRISO particles. The influences of the pretreatment methods such as heating oxidation of graphite, hydrothermal and oxidants oxidation were investigated in the present work. The experimental results showed that there were no significant effects on increasing the disintegration rate when pretreatment methods were used ahead of electrochemical disintegration. This phenomenon indicated that the fuel elements which were calcined at 1073 K and pressed under 300 MPa are too compact to be broken by these pretreatment methods. And the electrochemical disintegration is an effective but slow method in breaking the graphite matrix. (author)

  19. The commercial application prospect of HTGR plant in China

    International Nuclear Information System (INIS)

    Wang Yingsu

    2008-01-01

    With an introduction of the features and current situation of the HTGR power generation as well as the development of HTGR demonstration project in China, the article analyzes the necessity of developing HTGR power plants. The article proposes to exercise the advantages of HTGR to full extent so as to further develop HTGR power plants. It is believed that HTGR is of great commercial promotion value under appropriate circumstances. (authors)

  20. Process heat supply requirements on HTGRs

    International Nuclear Information System (INIS)

    Schad, M.K.

    1989-01-01

    Since it has been claimed that the MHTGR is competitive with coal in producing electricity, the MHTGR must be competitive in producing process heat. There is a huge process heat market and there are quite a number of processes where the industrial MHTGR = HTRI could supply the necessary process heat and energy. However, to enhance its introduction on the market and to conquer a reasonable share of the market, the HTRI should fulfill the following major requirements: Unlimited constant and flexible heat supply, no secondary heat transport system at higher temperatures and low radioactive contamination level of the primary helium. Unlimited constant and flexible heat supply could be achieved with smaller HTRIs having heat generation capacities below 100 MW-th. The process heat generated by smaller HTRIs need not be more expensive since the installed necessary heat supply redundancy is smaller and the excess power density lower. The process heat at elevated temperatures generated by a HTRI with a secondary heat transfer system is much more expensive due to the additional investment and operating cost as well as the reduced helium temperature span available. For some processes, the HTRI is not able to cover the total process heat requirement while other processes can consume only part of the heat offered. These limitations could be reduced by using higher core outlet and inlet temperatures or both. Due to the considerably lower heat transfer rates and the resulting larger heat transfer areas in process plants, the diffusion of nuclear activity at elevated temperatures may increase so that a more efficient helium cleaning system may be required. (author). 5 figs, 3 tabs

  1. Process heat recovery: hot prospects

    Energy Technology Data Exchange (ETDEWEB)

    1982-03-01

    By updating established technologies to recover heat at higher temperatures and under more corrosive conditions, British industry could recover six to eight million tons of coal equivalent that it currently wastes. Organic liquids in organic Rankine cycle (ORC) engines and simpler designs than steam turbines can increase efficiency. They also eliminate the need for vacuum pumps and permit the use of air cooling. Cooperative government-private industry research programs are exploring the use of ORC engines. Other heat-recovery projects include a Scottish paper mill, a metal decorating and printing plant, a falling-cloud heat exchanger, and heat-pipe development. 4 figures, 1 table. (DCK)

  2. Heating and cooling processes in disks*

    Directory of Open Access Journals (Sweden)

    Woitke Peter

    2015-01-01

    Full Text Available This chapter summarises current theoretical concepts and methods to determine the gas temperature structure in protoplanetary disks by balancing all relevant heating and cooling rates. The processes considered are non-LTE line heating/cooling based on the escape probability method, photo-ionisation heating and recombination cooling, free-free heating/cooling, dust thermal accommodation and high-energy heating processes such as X-ray and cosmic ray heating, dust photoelectric and PAH heating, a number of particular follow-up heating processes starting with the UV excitation of H2, and the release of binding energy in exothermal reactions. The resulting thermal structure of protoplanetary disks is described and discussed.

  3. Nuclear reactor plant for production process heat

    International Nuclear Information System (INIS)

    Weber, M.

    1979-01-01

    The high temperature reactor is suitable as a heat source for carrying out endothermal chemical processes. A heat exchanger is required for separating the reactor coolant gases and the process medium. The heat of the reactor is transferred at a temperature lower than the process temperature to a secondary gas and is compressed to give the required temperature. The compression energy is obtained from the same reactor. (RW) [de

  4. 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

  5. HTGR Measurements and Instrumentation Systems

    International Nuclear Information System (INIS)

    Ball, Sydney J.; Holcomb, David Eugene; Cetiner, Mustafa Sacit

    2012-01-01

    This report provides an integrated overview of measurements and instrumentation for near-term future high-temperature gas-cooled reactors (HTGRs). Instrumentation technology has undergone revolutionary improvements since the last HTGR was constructed in the United States. This report briefly describes the measurement and communications needs of HTGRs for normal operations, maintenance and inspection, fuel fabrication, and accident response. The report includes a description of modern communications technologies and also provides a potential instrumentation communications architecture designed for deployment at an HTGR. A principal focus for the report is describing new and emerging measurement technologies with high potential to improve operations, maintenance, and accident response for the next generation of HTGRs, known as modular HTGRs, which are designed with passive safety features. Special focus is devoted toward describing the failure modes of the measurement technologies and assessing the technology maturity.

  6. Graphite oxidation in HTGR atmosphere

    International Nuclear Information System (INIS)

    Growcock, F.B.; Barry, J.J.; Finfrock, C.C.; Rivera, E.; Heiser, J.H. III

    1982-01-01

    On-going and recently completed studies of the effect of thermal oxidation on the structural integrity of HTGR candidate graphites are described, and some results are presented and discussed. This work includes the study of graphite properties which may play decisive roles in the graphites' resistance to oxidation and fracture: pore size distribution, specific surface area and impurity distribution. Studies of strength loss mechanisms in addition to normal oxidation are described. Emphasis is placed on investigations of the gas permeability of HTGR graphites and the surface burnoff phenomenon observed during recent density profile measurements. The recently completed studies of catalytic pitting and the effects of prestress and stress on reactivity and ultimate strength are also discussed

  7. Examination on small-sized cogeneration HTGR for developing countries

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; Mozumi, Yasuhiro; Imai, Yoshiyuki; Tanaka, Nobuyuki; Okuda, Hiroyuki; Iwatsuki, Jin; Kubo, Shinji; Takada, Shoji; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2008-03-01

    The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors. (author)

  8. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  9. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  10. Waste management considerations in HTGR recycle operations

    International Nuclear Information System (INIS)

    Pence, D.T.; Shefcik, J.J.; Heath, C.A.

    1975-01-01

    Waste management considerations in the recycle of HTGR fuel are different from those encountered in the recycle of LWR fuel. The types of waste associated with HTGR recycle operations are discussed, and treatment methods for some of the wastes are described

  11. OMR type process heat reactor

    International Nuclear Information System (INIS)

    Franzetti, Franco.

    1974-01-01

    A description is given of an OMR type reactor for heat generation. It includes a vessel the upper part of which is shut by a plug. The lower part of the vessel includes a core of fuel elements and is filled with an organic liquid. Over this there is a middle area filled with an inert gas. The plug includes an upper part forming a closure and resting around its edge on the vessel, and a lower part fixed under the closure and composed of a hollow cylindrical tank fitted with a bottom and filled with another organic liquid. The height of the cylindrical tank is such that, increased by the height of the first organic liquid in the lower area and above the core, it provides biological protection. The cooling system includes a heat exchanger and a pump to move the liquid from the lower part of the core and to inject some as spray into that part of the vessel filled with the inert gas. When loading and unloading, after the reactor is shut down, the clear organic liquid contained in the plug is discharged into the reactor vessel in such a way that it does not mix with the opaque organic liquid already contained in the vessel, and in that the opaque organic liquid is emptied out [fr

  12. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  13. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  14. Solar process heat is becoming sexy

    Energy Technology Data Exchange (ETDEWEB)

    Morhart, Alexander

    2011-07-01

    Linear concentrating solar collectors for solar medium-temperature process heat: an exotic niche market has turned into a wide range of offers for commercial and private customers - and there is no end in sight to the technical developments. (orig.)

  15. Microwave heating processes involving carbon materials

    Energy Technology Data Exchange (ETDEWEB)

    Menendez, J.A.; Arenillas, A.; Fidalgo, B.; Fernandez, Y.; Zubizarreta, L.; Calvo, E.G.; Bermudez, J.M. [Instituto Nacional del Carbon, CSIC, Apartado 73, 33080 Oviedo (Spain)

    2010-01-15

    Carbon materials are, in general, very good absorbents of microwaves, i.e., they are easily heated by microwave radiation. This characteristic allows them to be transformed by microwave heating, giving rise to new carbons with tailored properties, to be used as microwave receptors, in order to heat other materials indirectly, or to act as a catalyst and microwave receptor in different heterogeneous reactions. In recent years, the number of processes that combine the use of carbons and microwave heating instead of other methods based on conventional heating has increased. In this paper some of the microwave-assisted processes in which carbon materials are produced, transformed or used in thermal treatments (generally, as microwave absorbers and catalysts) are reviewed and the main achievements of this technique are compared with those obtained by means of conventional (non microwave-assisted) methods in similar conditions. (author)

  16. Electromagnetic heating processes: analysis and simulations

    OpenAIRE

    Calay, Rajnish Kaur

    1994-01-01

    Electromagnetic heating (EMH) processes are being increasingly used in the industrial and domestic sectors, yet they receive relatively little attention in the thermal engineering domain. Time-temperature characteristics in EMH are qualitatively different from those in conventional heating techniques due to the additional parameters (viz dielectric properties of the material, size and shape of the product and process frequency). From a unified theory perspective, a multi-...

  17. Advances in Nuclear Power Process Heat Applications

    International Nuclear Information System (INIS)

    2012-05-01

    Following an IAEA coordinated research project, this publication compiles the findings of research and development activities related to practical nuclear process heat applications. An overview of current progress on high temperature gas cooled reactors coupling schemes for different process heat applications, such as hydrogen production and desalination is included. The associated safety aspects are also highlighted. The summary report documents the results and conclusions of the project.

  18. Thermodynamic analysis on theoretical models of cycle combined heat exchange process: The reversible heat exchange process

    International Nuclear Information System (INIS)

    Zhang, Chenghu; Li, Yaping

    2017-01-01

    Concept of reversible heat exchange process as the theoretical model of the cycle combined heat exchanger could be useful to determine thermodynamics characteristics and the limitation values in the isolated heat exchange system. In this study, the classification of the reversible heat exchange processes is presented, and with the numerical method, medium temperature variation tendency and the useful work production and usage in the whole process are investigated by the construction and solution of the mathematical descriptions. Various values of medium inlet temperatures and heat capacity ratio are considered to analyze the effects of process parameters on the outlet temperature lift/drop. The maximum process work transferred from the Carnot cycle region to the reverse cycle region is also researched. Moreover, influence of the separating point between different sub-processes on temperature variation profile and the process work production are analyzed. In addition, the heat-exchange-enhancement-factor is defined to study the enhancement effect of the application of the idealized process in the isolated heat exchange system, and the variation degree of this factor with process parameters change is obtained. The research results of this paper can be a theoretical guidance to construct the cycle combined heat exchange process in the practical system. - Highlights: • A theoretical model of Cycle combined heat exchange process is proposed. • The classification of reversible heat exchange process are presented. • Effects of Inlet temperatures and heat capacity ratio on process are analyzed. • Process work transmission through the whole process is studied. • Heat-exchange-enhancement-factor can be a criteria to express the application effect of the idealized process.

  19. Uncertainties in HTGR neutron-physical characteristics due to computational errors and technological tolerances

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Grebennik, V.N.; Davidenko, V.G.; Kosovskij, V.G.; Smirnov, O.N.; Tsibul'skij, V.F.

    1991-01-01

    The paper is dedicated to the consideration of uncertainties is neutron-physical characteristics (NPC) of high-temperature gas-cooled reactors (HTGR) with a core as spherical fuel element bed, which are caused by calculations from HTGR parameters mean values affecting NPC. Among NPC are: effective multiplication factor, burnup depth, reactivity effect, control element worth, distribution of neutrons and heat release over a reactor core, etc. The short description of calculated methods and codes used for HTGR calculations in the USSR is given and evaluations of NPC uncertainties of the methodical character are presented. Besides, the analysis of the effect technological deviations in parameters of reactor main elements such as uranium amount in the spherical fuel element, number of neutron-absorbing impurities in the reactor core and reflector, etc, upon the NPC is carried out. Results of some experimental studies of NPC of critical assemblies with graphite moderator are given as applied to HTGR. The comparison of calculations results and experiments on critical assemblies has made it possible to evaluate uncertainties of calculated description of HTGR NPC. (author). 8 refs, 8 figs, 6 tabs

  20. The desorption of caesium from Peach Bottom HTGR steam generator materials

    International Nuclear Information System (INIS)

    Clark, M.J.

    1979-03-01

    The work at Harwell on the Peach Bottom End-of-Life Program in co-operation with the General Atomic Company (U.S.A.) is described. Materials taken from the Economiser, Evaporator and Superheater Sections of the Peach Bottom Unit No. 1. High Temperature Gas Cooled Reactor (HTGR) Heat Exchanger were placed in a reducing atmosphere comparable to the composition of an HTGR helium coolant gas, and the desorption of caesium isotopes measured under known conditions of flow, temperature and oxygen pressure. (author)

  1. METAL CHIP HEATING PROCESS INVESTIGATION (Part I

    Directory of Open Access Journals (Sweden)

    O. M. Dyakonov

    2007-01-01

    Full Text Available The main calculation methods for heat- and mass transfer in porous heterogeneous medium have been considered. The paper gives an evaluation of the possibility to apply them for calculation of metal chip heating process. It has been shown that a description of transfer processes in a chip has its own specific character that is attributed to difference between thermal and physical properties of chip material and lubricant-coolant components on chip surfaces. It has been determined that the known expressions for effective heat transfer coefficients can be used as basic ones while approaching mutually penetrating continuums. A mathematical description of heat- and mass transfer in chip medium can be considered as a basis of mathematical modeling, numerical solution and parameter optimization of the mentioned processes.

  2. Relaxation processes during amorphous metal alloys heating

    International Nuclear Information System (INIS)

    Malinochka, E.Ya.; Durachenko, A.M.; Borisov, V.T.

    1982-01-01

    Behaviour of Te+15 at.%Ge and Fe+13 at.%P+7 at.%C amorphous metal alloys during heating has been studied using the method of differential scanning calorimetry (DSC) as the most convenient one for determination of the value of heat effects, activation energies, temperature ranges of relaxation processes. Thermal effects corresponding to high-temperature relaxation processes taking place during amorphous metal alloys (AMA) heating are detected. The change of ratio of relaxation peaks values on DSC curves as a result of AMA heat treatment can be explained by the presence of a number of levels of inner energy in amorphous system, separated with potential barriers, the heights of which correspond to certain activation energies of relaxation processes

  3. HTGR Gas Turbine Program. Semiannual progress report for the period ending September 30, 1979

    International Nuclear Information System (INIS)

    1980-05-01

    Information on the HTGR-GT program is presented concerning systems design methods; systems dynamics methods; alternate design; miscellaneous controls and auxiliary systems; structural mechanics; shielding analysis; licensing; safety; availability; reactor turbine system integration with plant; PCRV liners, penetrations, and closures; PCRV structures; thermal barrier; reactor internals; turbomachinery; turbomachine remote maintenance; control valve; heat exchangers; plant protection system; and plant control system

  4. New HTGR plant concept with inherently safe features aimed at small energy users needs

    International Nuclear Information System (INIS)

    McDonald, C.F.; Silady, F.S.; Shenoy, A.S.

    1982-01-01

    A small high-temperature gas-cooled reactor (HTGR) concept is proposed which could provide the energy needs for certain sectors of industrialized nations and the developing countries. The key to the economic success for small reactors, which have potential benefits for special markets, lies in altering the traditional scaling laws. Toward this goal, a small HTGR concept embodying passive decay heat removal features is currently being evaluated. This paper emphasizes the safety-related aspects of a small HTGR. The proposed small reactor concept is new and still in the design development stage, and a significant effort must be expended to establish a design which is technically and economically feasible and will meet the increasingly demanding safety and licensing goals for reactors of the future

  5. Heat transfer in a thermoacoustic process

    International Nuclear Information System (INIS)

    Beke, Tamas

    2012-01-01

    Thermoacoustic instability is defined as the excitation of acoustic modes in chambers with heat sources due to the coupling between acoustic perturbations and unsteady heat addition. The major objective of this paper is to achieve accurate theoretical results in a thermoacoustic heat transfer process. We carry out a detailed heat transfer analysis aimed at determining the stability–instability border of the thermoacoustic system. In this paper, we present a project type of physical examination and modelling task. We employed an electrically heated Rijke tube in our thermoacoustic project work. The aim of our project is to help our students enlarge their knowledge about thermodynamics, mainly about thermoacoustics, and develop their applied information technology and mathematical skills. (paper)

  6. INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

    Directory of Open Access Journals (Sweden)

    Sudarmono Sudarmono

    2015-03-01

    Full Text Available The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR. Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology  (PTRKN has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generation system, which has high energy efficiency, and has high and clean inherent safety level. The geometry and structure of the HTGR200 core are designed to produce the output of helium gas coolant temperature as high as 950 °C to be used for hydrogen production and other industrial processes in co-generative way. The output of very high temperature helium gas will cause thermal stress on the fuel pebble that threats the integrity of fission product confinement. Therefore, it is necessary to perform thermal-flow evaluation to determine the temperature distribution in the graphite and fuel pebble in the HTGR core. The evaluation was carried out by Thermix-Konvek module code that has been already integrated into VSOP'94 code. The HTGR core geometry was done using BIRGIT module code for 2-D model (RZ model with 5 channels of pebble flow in active core in the radial direction. The evaluation results showed that the highest and lowest temperatures in the reactor core are 999.3 °C and 886.5 °C, while the highest temperature of TRISO UO2 is 1510.20 °C in the position (z= 335.51 cm; r=0 cm. The analysis done based on reactor condition of 120 kg/s of coolant mass flow rate, 7 MPa of pressure and 200 MWth of power. Compared to the temperature distribution resulted between VSOP’94 code and fuel temperature limitation as high as 1600 oC, there is enough safety margin from melting or disintegrating. Keywords: Thermal-Flow, VSOP’94, Thermix-Konvek, HTGR, temperature

  7. Generator technology for HTGR power plants

    International Nuclear Information System (INIS)

    Lomba, D.; Thiot, D.

    1997-01-01

    Approximately 15% of the worlds installed capacity in electric energy production is from generators developed and manufactured by GEC Alsthom. GEC Alsthom is now working on the application of generators for HTGR power conversion systems. The main generator characteristics induced by the different HTGR power conversion technology include helium immersion, high helium pressure, brushless excitation system, magnetic bearings, vertical lineshaft, high reliability and long periods between maintenance. (author)

  8. Improving Process Heating System Performance v3

    Energy Technology Data Exchange (ETDEWEB)

    None

    2016-04-11

    Improving Process Heating System Performance: A Sourcebook for Industry is a development of the U.S. Department of Energy (DOE) Advanced Manufacturing Office (AMO) and the Industrial Heating Equipment Association (IHEA). The AMO and IHEA undertook this project as part of an series of sourcebook publications developed by AMO on energy-consuming industrial systems, and opportunities to improve performance. Other topics in this series include compressed air systems, pumping systems, fan systems, steam systems, and motors and drives

  9. High temperature nuclear process heat systems for chemical processes

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.

    1976-01-01

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

  10. Oil shales and the nuclear process heat

    International Nuclear Information System (INIS)

    Scarpinella, C.A.

    1974-01-01

    Two of the primary energy sources most dited as alternatives to the traditional fossil fuels are oil shales and nuclear energy. Several proposed processes for the extraction and utilization of oil and gas from shale are given. Possible efficient ways in which nuclear heat may be used in these processes are discussed [pt

  11. Proceedings of the 2nd JAERI symposium on HTGR technologies October 21 ∼ 23, 1992, Oarai, Japan

    International Nuclear Information System (INIS)

    1993-01-01

    The Japan Atomic Energy Research Institute (JAERI) held the 2nd JAERI Symposium on HTGR Technologies on October 21 to 23, 1992, at Oarai Park Hotel at Oarai-machi, Ibaraki-ken, Japan, with support of International Atomic Energy Agency (IAEA), Science and Technology Agency of Japan and the Atomic Energy Society of Japan on the occasion that the construction of the High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, is now being proceeded smoothly. In this symposium, the worldwide present status of research and development (R and D) of the HTGRs and the future perspectives of the HTGR development were discussed with 47 papers including 3 invited lectures, focusing on the present status of HTGR projects and perspectives of HTGR Development, Safety, Operation Experience, Fuel and Heat Utilization. A panel discussion was also organized on how the HTGRs can contribute to the preservation of global environment. About 280 participants attended the symposium from Japan, Bangladesh, Germany, France, Indonesia, People's Republic of China, Poland, Russia, Switzerland, United Kingdom, United States of America, Venezuela and the IAEA. This paper was edited as the proceedings of the 2nd JAERI Symposium on HTGR Technologies, collecting the 47 papers presented in the oral and poster sessions along with 11 panel exhibitions on the results of research and development associated to the HTTR. (author)

  12. 1170-MW(t) HTGR-PS/C plant application study report: Geismar, Louisiana refinery/chemical complex application

    International Nuclear Information System (INIS)

    McMain, A.T. Jr.; Stanley, J.D.

    1981-05-01

    This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to an industrial complex at Geismar, Louisiana. This study compares the HTGR with coal and oil as process plant fuels. This study uses a previous broad energy alternative study by the Stone and Webster Corporation on refinery and chemical plant needs in the Gulf States Utilities service area. The HTGR-PS/C was developed by General Atomic (GA) specifically for industries which require both steam and electric energy. The GA 1170-MW(t) HTGR-PC/C design is particularly well suited to industrial applications and is expected to have excellent cost benefits over other energy sources

  13. HTGR development in the United States of America

    International Nuclear Information System (INIS)

    Fox, J.E.

    1991-01-01

    The status of high temperature gas-cooled reactors (HTGR) development in the United States of America is described, including the organizational structure for the development support, HTGR development programme, and plans for future activities in the field

  14. Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors

    International Nuclear Information System (INIS)

    Sabharwall, Piyush; Kim, Eung Soo; Anderson, Nolan

    2011-01-01

    The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

  15. Present status of research on hydrogen energy and perspective of HTGR hydrogen production system

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Ogawa, Masuro; Akino, Norio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-03-01

    A study was performed to make a clear positioning of research and development on hydrogen production systems with a High Temperature Gas-cooled Reactor (HTGR) under currently promoting at the Japan Atomic Energy Research Institute through a grasp of the present status of hydrogen energy, focussing on its production and utilization as an energy in future. The study made clear that introduction of safe distance concept for hydrogen fire and explosion was practicable for a HTGR hydrogen production system, including hydrogen properties and need to provide regulations applying to handle hydrogen. And also generalization of hydrogen production processes showed technical issues of the HTGR system. Hydrogen with HTGR was competitive to one with fossil fired system due to evaluation of production cost. Hydrogen is expected to be used as promising fuel of fuel cell cars in future. In addition, the study indicated that there were a large amount of energy demand alternative to high efficiency power generation and fossil fuel with nuclear energy through the structure of energy demand and supply in Japan. Assuming that hydrogen with HTGR meets all demand of fuel cell cars, an estimation would show introduction of the maximum number of about 30 HTGRs with capacity of 100 MWt from 2020 to 2030. (author)

  16. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

  17. Proceedings of the 1st JAEA/KAERI information exchange meeting on HTGR and nuclear hydrogen technology

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Sakaba, Nariaki; Nishihara, Tetsuo; Yan, Xing L.; Hino, Ryutaro

    2007-03-01

    Japan Atomic Energy Agency (JAEA) has completed an implementation with Korea Atomic Energy Research Institute (KAERI) on HTGR and nuclear hydrogen technology, 'The Implementation of Cooperative Program in the Field of Peaceful Uses of Nuclear Energy between KAERI and JAEA. 'To facilitate efficient technology development on HTGR and nuclear hydrogen by the IS process, an information exchange meeting was held at the Oarai Research and Development Center of JAEA on August 28-30, 2006 under Program 13th of the JAEA/KAERI Implementation, 'Development of HTGR and Nuclear Hydrogen Technology'. JAEA and KAERI mutually showed the status and future plan of the HTTR (High-Temperature Engineering Test Reactor) project in Japan and of the NHDD (Nuclear Hydrogen Development and Demonstration) project in Korea, respectively, and discussed collaboration items. This proceedings summarizes all materials of presented technical discussions on HTGR and hydrogen production technology as well as the meeting briefing including collaboration items. (author)

  18. Modeling of Dielectric Heating within Lyophilization Process

    Directory of Open Access Journals (Sweden)

    Jan Kyncl

    2014-01-01

    Full Text Available A process of lyophilization of paper books is modeled. The process of drying is controlled by a dielectric heating system. From the physical viewpoint, the task represents a 2D coupled problem described by two partial differential equations for the electric and temperature fields. The material parameters are supposed to be temperature-dependent functions. The continuous mathematical model is solved numerically. The methodology is illustrated with some examples whose results are discussed.

  19. Air ingress behavior during a primary-pipe rupture accident of HTGR

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki

    1997-11-01

    The inherent properties of a HTGR facilitates the design with high degree of passive safe performances, compared to other type. However, it is still not clear if the present HTGR can maintain a passive safe function during a primary-pipe rupture accident, or what would be design criteria to guarantee the HTGR with the high degree of passive safe performances during the accident. To investigate safe characteristics, the study has been performed experimentally and analytically on the air ingress behavior during the accident. It was indicated that there are two stages in the accident of the HTGR having a reverse U-shaped channel. In the first stage, an air ingress process limits molecular diffusion and natural circulation of the gas mixture having a very slow velocity. In the second stage, the air ingress process limits the ordinary natural circulation of air throughout the reactor. A numerical calculation code has been developed to analyze thermal-hydraulic behavior during the first stage. This code provides a numerical method for analyzing a transport phenomena in a multi-component gas system by solving one-dimensional basic equations and using a flow network model. It was possible to predict or analyze the air ingress process regarding the density of the gas mixture, concentration of each gas species and duration of the first stage of the accident. It was indicated that the safe characteristics of the HTGR from the present experiment as follows. The safety cooling rate that the air ingress process terminates during the first stage exists in the HTGR having the reverse U-shaped channel. Moreover, the ordinary natural circulation of air can not produce in the second stage by injecting helium from the bottom of the pressure vessel corresponding the low-temperature side channel. Therefore, it was found that the idea of helium injection is one of useful methods for the prevention of air ingress and of graphite corrosion in the future HTGRs. (J.P.N.). 74 refs

  20. Thermal stress analysis of HTGR fuel and control rod fuel blocks in the HTGR in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new approach that utilizes the equivalent solid plate method has been applied to the thermal stress analysis of HTGR fuel and control rod fuel blocks. Cases were considered where these blocks, loaded with reprocessed HTGR fuel pellets, were being cured at temperatures up to 1800 0 C. A two-dimensional segment of a fuel block cross section including fuel, coolant holes, and graphite matrix was analyzed using the ORNL HEATING3 heat transfer code to determine the temperature-dependent effective thermal conductivity for the perforated region of the block. Using this equivalent conductivity to calculate the temperature distributions through different cross sections of the blocks, two-dimensional thermal-stress analyses were performed through application of the equivalent solid plate method. In this approach, the perforated material is replaced by solid homogeneous material of the same external dimensions but whose material properties have been modified to account for the perforations

  1. Laser Processed Condensing Heat Exchanger Technology Development

    Science.gov (United States)

    Hansen, Scott; Wright, Sarah; Wallace, Sarah; Hamilton, Tanner; Dennis, Alexander; Zuhlke, Craig; Roth, Nick; Sanders, John

    2017-01-01

    The reliance on non-permanent coatings in Condensing Heat Exchanger (CHX) designs is a significant technical issue to be solved before long-duration spaceflight can occur. Therefore, high reliability CHXs have been identified by the Evolvable Mars Campaign (EMC) as critical technologies needed to move beyond low earth orbit. The Laser Processed Condensing Heat Exchanger project aims to solve these problems through the use of femtosecond laser processed surfaces, which have unique wetting properties and potentially exhibit anti-microbial growth properties. These surfaces were investigated to identify if they would be suitable candidates for a replacement CHX surface. Among the areas researched in this project include microbial growth testing, siloxane flow testing in which laser processed surfaces were exposed to siloxanes in an air stream, and manufacturability.

  2. Numerical simulation of heat transfer process in automotive brakes

    OpenAIRE

    Gonzalo Voltas, David

    2013-01-01

    This master thesis concerns the theoretical investigations of the heat transfer process in automotive brakes. The process of heat generation and heat transfer to ambient air in automotive brake was presented. The two–dimensional, axi-symmetrical model of transient heat conduction for the brake was applied. The relevant boundary conditions, that describe the heat generated in the brake and the heat transferred to ambient air, were used. The unsteady heat conduction problem was solved by the...

  3. Preliminary risk assessments of the small HTGR

    International Nuclear Information System (INIS)

    Everline, C.J.; Bellis, E.A.

    1985-05-01

    Preliminary investment and safety risk assessments were performed for a preconceptual design of a four-module 250-MW(t) side-by-side steel-vessel pebble bed HTGR plant. Broad event spectra were analyzed involving component damage resulting in unscheduled plant outages and fission product releases resulting in offsite doses. The preliminary assessment indicates at this stage of the design that two categories of events govern the investment risk envelope: primary coolant leaks which release some circulating and plate-out activity that contaminates the confinement and turbogenerator damage which involves extensive turbine blade failure. Primary coolant leaks are important contributors because associated cleanup and decontamination requirements result in longer outages that arise from other events with comparable frequencies. Turbogenerator damage is the salient low-frequency investment risk accident due to the relatively long outages being experienced in the industry. Thermal transients are unimportant investment risk contributors because pressurized core heatups cause little damage, and depressurized core heatups occur at negligible frequencies relative to the forced outage goal. These preliminary results demonstrate investment and safety risk goal compliance at this stage in the design process. Studies are continuing in order to provide valuable insights into risk-significant events to assure a balanced approach to meeting user and regulatory requirements

  4. HTR's role in process heat applications

    International Nuclear Information System (INIS)

    Kuhr, Reiner

    2008-01-01

    Advanced high-temperature nuclear reactors create a number of new opportunities for nuclear process heat applications. These opportunities are based on the high-temperature heat available, smaller reactor sizes, and enhanced safety features that allow siting close to process plants. Major sources of value include the displacement of premium fuels and the elimination of CO 2 emissions from combustion of conventional fuels and their use to produce hydrogen. High value applications include steam production and cogeneration, steam methane reforming, and water splitting. Market entry by advanced high-temperature reactor technology is challenged by the evolution of nuclear licensing requirements in countries targeted for early applications, by the development of a customer base not familiar with nuclear technology and related issues, by convergence of oil industry and nuclear industry risk management, by development of public and government policy support, by resolution of nuclear waste and proliferation concerns, and by the development of new business entities and business models to support commercialization. New HTR designs may see a larger opportunity in process heat niche applications than in power given competition from larger advanced light water reactors. Technology development is required in many areas to enable these new applications, including the commercialization of new heat exchangers capable of operating at high temperatures and pressures, convective process reactors and suitable catalysts, water splitting system and component designs, and other process-side requirements. Key forces that will shape these markets include future fuel availability and pricing, implementation and monetization of CO 2 emission limits, and the formation of international energy and environmental policy that will support initiatives to provide the nuclear licensing frameworks and risk distribution needed to support private investment. This paper was developed based on a plenary

  5. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

    International Nuclear Information System (INIS)

    1981-05-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan

  6. HTGR Fuel-Technology Program. Semiannual report for the period ending September 30, 1982

    International Nuclear Information System (INIS)

    1982-11-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at GA Technologies Inc. during the second half of FY-1982. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was completed during this period was the preparation of input data for a long-range technology program plan

  7. Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

    International Nuclear Information System (INIS)

    Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

    1983-05-01

    This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

  8. Heat pump processes induced by laser radiation

    Science.gov (United States)

    Garbuny, M.; Henningsen, T.

    1980-01-01

    A carbon dioxide laser system was constructed for the demonstration of heat pump processes induced by laser radiation. The system consisted of a frequency doubling stage, a gas reaction cell with its vacuum and high purity gas supply system, and provisions to measure the temperature changes by pressure, or alternatively, by density changes. The theoretical considerations for the choice of designs and components are dicussed.

  9. HTGR fuel particle crusher design evaluation

    International Nuclear Information System (INIS)

    Johanson, N.W.

    1978-10-01

    This report describes an evaluation of the design of the existing engineering-scale fuel particle crushing system for the HTGR reprocessing cold pilot plant at General Atomic Company (GA). The purpose of this evaluation is to assess the suitability of the existing design as a prototype of the HTGR Recycle Reference Facility (HRRF) particle crushing system and to recommend alternatives where the existing design is thought to be unsuitable as a prototype. This evaluation has led to recommendations for an upgraded design incorporating improvements in bearing and seal arrangement, housing construction, and control of roll gap thermal expansion. 23 figures, 6 tables

  10. The prospects of HTGR in China

    International Nuclear Information System (INIS)

    Sun, Y.; Tong, Y.; Wu, Z.

    1994-01-01

    Present situations of the energy market in China are briefly introduced, while the forecast of the possible development of the Chinese energy market is shortly discussed. The discussion focuses on the expected roles of high temperature gas-cooled reactors (HTGR) in the Chinese energy market in the next century. The history and present status of the development of HTGR technologies in China are presented. In the National High-Tech Programme, a 10 MW helium-cooled test reactor (HTR-10) is projected to be built within this century. The main technical and safety features of the HTR-10 reactor are discussed. (author)

  11. Calcination, Reduction and Sintering of ADU Spheres for HTGR Fuel

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Eom, Sung Ho; Kim, Yeon Ku; Kim, Woong Ki; Kim, Young Min; Lee, Young Woo; Kim, Ju Hee; Cho, Hyo Jin; Cho, Moon Seoung

    2011-01-01

    The international oil market is again in turmoil in accordance with the increasing of human needs and energy consumption. Soaring oil prices, fears of supply security, and climate change are concerned becoming more concrete make for an uncertain energy future. In this view point, nuclear energy is an important, yet controversial option for energy supply. High Temperature Gas Reactor will play a dominant role in the worldwide fleet of nuclear reactors of the next decade for electricity production and high temperature heat. HTGR have two reactor types which use the basic fuel concept based on the dispersion of TRISO coated particles in graphite in shown Fig.1. The TRISO coated particle for these purposes is prepared with pyro-carbon and silicone carbide coatings on a spherical UO 2 kernel surface as fissile material. The TRISO fuel particle consists of a microsphere (i.e., UO 2 kernel) of nuclear material: encapsulated by multiple layers of pyro-carbon and a SiC layer. This multiple coating layers system has been engineered to retain the fission products generated by fission of the nuclear material in the kernel during normal operation and all licensing basis events over the design lifetime of the fuel. UO 2 kernels are produced by using the modified sol-gel process, a wet process, generally known as the GSP method. Wet chemical processes are flexible in producing kernels of different size and chemical composition with high throughout and yield, good spherical shape, and narrow size distribution. This chemical processing route is well-known to the potential kernel fabrication processes. The principle, as set out in Fig.2, involves first of all preparing a pseudo-sol(also known as a 'broth') from an initial uranyl nitrate solution . This broth solution is obtained through addition of a number of additives, as determined by process know-how, including a soluble organic polymer, that are subsequently gels into droplets and are dispersed for ADU precipitation. The

  12. Potential of the HTGR hydrogen cogeneration system in Japan

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Mouri, Tomoaki; Kunitomi, Kazuhiko

    2007-01-01

    A high temperature gas cooled reactor (HTGR) is one of the next generation nuclear systems. The HTGR hydrogen cogeneration system can produce not only electricity but also hydrogen. Then it has a potential to supply massive low-cost hydrogen without greenhouse gas emission for the future hydrogen society. Japan Atomic Energy Agency (JAEA) has been carried out the design study of the HTGR hydrogen cogeneration system (GTHTR300C). The thermal power of the reactor is 600 MW. The hydrogen production plant utilizes 370 MW and can supply 52,000 m 3 /h (0.4 Bm 3 /y) of hydrogen. Present industrial hydrogen production capacity in Japan is about 18 Bm 3 /y and it will decrease by 15 Bm 3 /y in 2030 due to the aging facilities. On the other hand, the hydrogen demand for fuel cell vehicle (FCV) in 2030 is estimated at 15 Bm 3 /y at a maximum. Since the hydrogen supply may be short after 2030, the additional hydrogen should be produced by clean hydrogen process to reduce greenhouse gas emission. This hydrogen shortage is a potential market for the GTHTR300C. The hydrogen production cost of GTHTR300C is estimated at 20.5 JPY/Nm 3 which has an economic competitiveness against other industrial hydrogen production processes. 38 units of the GTHTR300C can supply a half of this shortage which accounts for the 33% of hydrogen demand for FCV in 2100. According to the increase of hydrogen demand, the GTHTR300C should be constructed after 2030. (author)

  13. Status of reprocessing technology in the HTGR fuel cycle

    International Nuclear Information System (INIS)

    Kaiser, G.; Merz, E.; Zimmer, E.

    1977-01-01

    For more than ten years extensive R and D work has been carried out in the Federal Republic of Germany in order to develop the technology necessary for closing the fuel cycle of high-temperature gas-cooled reactors. The efforts are concentrated primarily on fuel elements having either highly enriched 235 U or recycled 233 U as the fissile and thorium as the fertile material embedded in a graphite matrix. They include the development of processes and equipment for reprocessing and remote preparation of coated microspheres from the recovered uranium. The paper reviews the issues and problems associated with the requirements to deal with high burn-up fuel from HTGR's of different design and composition. It is anticipated that a grind-burn-leach head-end treatment and a modified THOREX-type chemical processing are the optimum choice for the flowsheet. An overview of the present status achieved in construction of a small reprocessing facility, called JUPITER, is presented. It includes a discussion of problems which have already been solved and which have still to be solved like the treatment of feed/breed particle systems and for minimizing environmental impacts envisaged with a HTGR fuel cycle technology. Also discussed is the present status of remote fuel kernel fabrication and coating technology. Additional activities include the design of a mock-up prototype burning head-end facility, called VENUS, with a throughput equivalent to about 6000 MW installed electrical power, as well as a preliminary study for the utilisation of the Karlsruhe LWR prototype reprocessing plant (WAK) to handle HTGR fuel after remodelling of the installations. The paper concludes with an outlook of projects for the future

  14. Design of common heat exchanger network for batch processes

    International Nuclear Information System (INIS)

    Anastasovski, Aleksandar

    2014-01-01

    Heat integration of energy streams is very important for the efficient energy recovery in production systems. Pinch technology is a very useful tool for heat integration and maximizing energy efficiency. Creating of heat exchangers network as a common solution for systems in batch mode that will be applicable in all existing time slices is very difficult. This paper suggests a new methodology for design of common heat exchanger network for batch processes. Heat exchanger network designs were created for all determined repeatable and non-repeatable time periods – time slices. They are the basis for creating the common heat exchanger network. The common heat exchanger network as solution, satisfies all heat-transfer needs for each time period and for every existing combination of selected streams in the production process. This methodology use split of some heat exchangers into two or more heat exchange units or heat exchange zones. The reason for that is the multipurpose use of heat exchangers between different pairs of streams in different time periods. Splitting of large heat exchangers would maximize the total heat transfer usage of heat exchange units. Final solution contains heat exchangers with the minimum heat load as well as the minimum need of heat transfer area. The solution is applicable for all determined time periods and all existing stream combinations. - Highlights: •Methodology for design of energy efficient systems in batch processes. •Common Heat Exchanger Network solution based on designs with Pinch technology. •Multipurpose use of heat exchangers in batch processes

  15. In situ heat treatment process utilizing a closed loop heating system

    Science.gov (United States)

    Vinegar, Harold J.; Nguyen, Scott Vinh

    2010-12-07

    Systems and methods for an in situ heat treatment process that utilizes a circulation system to heat one or more treatment areas are described herein. The circulation system may use a heated liquid heat transfer fluid that passes through piping in the formation to transfer heat to the formation. In some embodiments, the piping may be positioned in at least two of the wellbores.

  16. Intermediate heat exchanger for HTR process heat application

    International Nuclear Information System (INIS)

    Crambes, M.

    1980-01-01

    In the French study on the nuclear gasification of coal, the following options were recommended: Coal hydrogenation, the hydrogen being derived from CH 4 reforming under the effects of HTR heat; the use of an intermediate helium circuit between the nuclear plant and the reforming plant. The purpose of the present paper is to describe the heat exchanger designed to transfer heat from the primary to the intermediate circuit

  17. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  18. Regulatory Framework of Safety for HTGR

    International Nuclear Information System (INIS)

    Huh, Chang Wook; Suh, Nam Duk

    2011-01-01

    Recent accident in Fukushima Daiichi plant in Japan makes big impacts on the future of nuclear business. Many countries are changing their nuclear projects and increased safety of nuclear plants is asked for from the public. Without providing safety the society accepts, it might be almost impossible to build new plants further. In this sense high temperature gas-cooled reactor (HTGR) which is under development needs to be licensed reflecting this new expectation regarding safety. It means we should have higher level of safety goal and a systematic regulatory framework to assure the safety. In our previous paper, we evaluated the current safety goal and design practice in view of this new safety expectation after Fukushima accident. It was argued that a top-down approach starting from safety goal is necessary to develop safety requirements or to assure safety. Thus we need to propose an ultimate safety goal public accepts and then establish a systematic regulatory framework. In this paper we are going to provide a conceptual regulatory framework to guarantee the safety of HTGR. Section 2 discusses the recent trend of IAEA safety requirements and then summarize the HTGR design approach. Incorporating these discussions, we propose a conceptual framework of regulation for safety of HTGR

  19. HTGR gas turbine power plant preliminary design

    International Nuclear Information System (INIS)

    Koutz, S.L.; Krase, J.M.; Meyer, L.

    1973-01-01

    The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references

  20. INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

    OpenAIRE

    Sudarmono Sudarmono

    2015-01-01

    The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR). Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology  (PTRKN) has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generati...

  1. Intensification of Evaporation and Condensation Processes in Heat Exchange Apparatus

    Directory of Open Access Journals (Sweden)

    L. L. Vasiliev

    2005-01-01

    Full Text Available The paper describes proposed design solutions for an intensification of heat transfer in evaporation and condensation heat exchangers. Complex experimental research of heat and mass transfer processes in flat and round cross-section miniature heat pipes is carried out. Optimization, development, manufacturing and an experimental investigation of copper miniature heat pipes with sintered powder are executed. Investigation results of capillary-porous structure properties that are used in evaporation and condensation heat-exchange apparatus are presented.

  2. Survey on the activities in Switzerland in the field of HTGR-development

    International Nuclear Information System (INIS)

    Sarlos, G.; Brogli, R.; Mathews, D.; Bucher, K.H.; Helbling, W.

    1991-01-01

    The activities of the Swiss industry and of the ''Paul Scherrer Institute'' in the development and production of components and systems for the nuclear industry are reviewed. For the HTGR, major programs include the German HTR-500 project, the gas-cooled district heating reactor (GHR), and the PROTEUS critical experiments. The experiments are being performed in the framework of an IAEA coordinated research program. (author)

  3. Thermal design and analysis of the HTGR fuel element vertical carbonizing and annealing furnace

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1977-06-01

    Computer analyses of the thermal design for the proposed HTGR fuel element vertical carbonizing and annealing furnace were performed to verify its capability and to determine the required power input and distribution. Although the furnace is designed for continuous operation, steady-state temperature distributions were obtained by assuming internal heat generation in the fuel elements to simulate their mass movement. The furnace thermal design, the analysis methods, and the results are discussed herein

  4. Heat source model for welding process

    International Nuclear Information System (INIS)

    Doan, D.D.

    2006-10-01

    One of the major industrial stakes of the welding simulation relates to the control of mechanical effects of the process (residual stress, distortions, fatigue strength... ). These effects are directly dependent on the temperature evolutions imposed during the welding process. To model this thermal loading, an original method is proposed instead of the usual methods like equivalent heat source approach or multi-physical approach. This method is based on the estimation of the weld pool shape together with the heat flux crossing the liquid/solid interface, from experimental data measured in the solid part. Its originality consists in solving an inverse Stefan problem specific to the welding process, and it is shown how to estimate the parameters of the weld pool shape. To solve the heat transfer problem, the interface liquid/solid is modeled by a Bezier curve ( 2-D) or a Bezier surface (3-D). This approach is well adapted to a wide diversity of weld pool shapes met for the majority of the current welding processes (TIG, MlG-MAG, Laser, FE, Hybrid). The number of parameters to be estimated is weak enough, according to the cases considered from 2 to 5 in 20 and 7 to 16 in 3D. A sensitivity study leads to specify the location of the sensors, their number and the set of measurements required to a good estimate. The application of the method on test results of welding TIG on thin stainless steel sheets in emerging and not emerging configurations, shows that only one measurement point is enough to estimate the various weld pool shapes in 20, and two points in 3D, whatever the penetration is full or not. In the last part of the work, a methodology is developed for the transient analysis. It is based on the Duvaut's transformation which overpasses the discontinuity of the liquid metal interface and therefore gives a continuous variable for the all spatial domain. Moreover, it allows to work on a fixed mesh grid and the new inverse problem is equivalent to identify a source

  5. Construction of the HTTR and its testing program for advanced HTGR development

    International Nuclear Information System (INIS)

    Tanaka, T.; Baba, O.; Shiozawa, S.; Okubo, M.; Kunitomi, K.

    1996-01-01

    Concerning about global warming due to emission of greenhouse effect gas like CO 2 , it is essentially important to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of CO 2 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) which can supply high-temperature heat with high thermal efficiency as well as high heat-utilizing efficiency. 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 950 deg. C 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). 2 refs, 2 figs, 1 tab., 2 photos

  6. HTGR-steam cycle/cogeneration plant economic potential

    International Nuclear Information System (INIS)

    1981-05-01

    The cogeneration of heat and electricity provides the potential for improved fuel utilization and corresponding reductions in energy costs. In the evaluation of the cogeneration plant product costs, it is advantageous to develop joint-product cost curves for alternative cogeneration plant models. The advantages and incentives for cogeneration are then presented in a form most useful to evaluate the various energy options. The HTGR-Steam Cycle/Cogeneration (SC/C) system is envisioned to have strong cogeneration potential due to its high-quality steam capability, its perceived nuclear siting advantages, and its projected cost advantages relative to coal. The economic information presented is based upon capital costs developed during 1980 and the economic assumptions identified herein

  7. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    Hilborn, J.S.; Seddon, W.A.; Barnstaple, A.G.

    1980-08-01

    It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

  8. Heat pipe cooling of power processing magnetics

    Science.gov (United States)

    Hansen, I. G.; Chester, M.

    1979-01-01

    The constant demand for increased power and reduced mass has raised the internal temperature of conventionally cooled power magnetics toward the upper limit of acceptability. The conflicting demands of electrical isolation, mechanical integrity, and thermal conductivity preclude significant further advancements using conventional approaches. However, the size and mass of multikilowatt power processing systems may be further reduced by the incorporation of heat pipe cooling directly into the power magnetics. Additionally, by maintaining lower more constant temperatures, the life and reliability of the magnetic devices will be improved. A heat pipe cooled transformer and input filter have been developed for the 2.4 kW beam supply of a 30-cm ion thruster system. This development yielded a mass reduction of 40% (1.76 kg) and lower mean winding temperature (20 C lower). While these improvements are significant, preliminary designs predict even greater benefits to be realized at higher power. This paper presents the design details along with the results of thermal vacuum operation and the component performance in a 3 kW breadboard power processor.

  9. Induction Heating Process Design Using COMSOL Multiphysics Software

    Directory of Open Access Journals (Sweden)

    Andy Triwinarko

    2011-08-01

    Full Text Available Induction heating is clean environmental heating process due to a non-contact heating process. There is lots of the induction heating type that be used in the home appliance but it is still new technology in Indonesia. The main interesting area of the induction heating design is the efficiency of the usage of energy and choice of the plate material. COMSOL Multiphysics Software can be used to simulate and estimate the induction heating process. Therefore, the software can be used to design the induction heating process that will have a optimum efficiency. The properties of the induction heating design were also simulated and analyzed such as effect of inductors width, inductors distance, and conductive plate material. The result was shown that the good design of induction heating must have a short width and distance inductor and used silicon carbide as material plate with high frequency controller.

  10. 1170-MW(t) HTGR-PS/C plant application study report: heavy oil recovery application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    This report describes the application of a high-temperature gas-cooled reactor (HTGR) which operates in a process steam/cogeneration (PS/C) mode in supplying steam for enhanced recovery of heavy oil and in exporting electricity. The technical and economic merits of an 1170-MW(t) HTGR-PS/C are compared with those of coal-fired plants and (product) oil-fired boilers for this application. The utility requirements for enhanced oil recovery were calculated by establishing a typical pattern of injection wells and production wells for an oil field similar to that of Kern County, California. The safety and licensing issues of the nuclear plant were reviewed, and a comparative assessment of the alternative energy sources was performed. Technically and economically, the HTGR-PS/C plant has attractive merits. The major offsetting factors would be a large-scale development of a heavy oil field by a potential user for the deployment of a 1170-MW(t) HTGR-PS/C; plant and the likelihood of available prime heavy oil fields for the mid-1990 operation

  11. Safety studies on heat transport and afterheat removal for GCR accident conditions

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1996-01-01

    The IAEA coordinated an international research program on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions (CRP-3)'. America, China, France, Germany, Japan, Netherlands and Russia participate the program. Final goal of the program is to show clearly to the world one of the most important salient features of the HTGR, that is the HTGR reactor can be cooled down by passive measures without causing any damage to the nuclear reactor system even in accidental conditions, and to make clear the boundaries (or restrictions) for the passive cooling regime. The first 5 year term of the coordinate program started in 1993 and established a goal to improve common knowledge for decay heat removal and to improve our tools, like computer codes and analytical models for the prediction of the performance of decay heat removal system. We are now performing benchmark problems for these purposes. The present efforts are concentrated on the benchmark for the passive heat removal performance outside the reactor vessel, partly because we have two different type of the HTGR in the world, the pebble bed type and the block type reactor. They have quite different heat dissipation behavior inside the reactor vessel. However, they have quite similar residual heat removal process outside the reactor vessel. For the first step of the international cooperation, we selected the common problem. After finishing the present benchmark we are planning to proceed to tackle the inside heat removal problem. (J.P.N.)

  12. 1170-MW(t) HTGR-PS/C plant application-study report: alumina-plant application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.; Stanley, J.D.

    1981-05-01

    This report considers the HTGR-PS/C application to producing alumina from bauxite. For the size alumina plant considered, the 1170-MW(t) HTGR-PS/C supplies 100% of the process steam and electrical power requirements and produces surplus electrical power and/or process steam, which can be used for other process users or electrical power production. Presently, the bauxite ore is reduced to alumina in plants geographically separated from the electrolysis plant. The electrolysis plants are located near economical electric power sources. However, with the integration of an 1170-MW(t) HTGR-PS/C unit in a commercial alumina plant, the excess electric power available [approx. 233 MW(e)] could be used for alumina electrolysis

  13. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  14. Safety criteria for advanced HTGR concepts

    International Nuclear Information System (INIS)

    Kroeger, W.

    1989-01-01

    It is commonly agreed that advanced HTGR concepts must be licensable, which means that they must fulfil existing regulatory requirements. Furthermore, it is necessary to improve their public acceptance and they must even be suitable for urban sites. Therefore, they should be 'safer' than existing plants, which mainly means with respect to low-frequency or beyond-design severe accidents. Last but not least, the realization of advanced HTGR would be easier if commonly shared safety principles could be stated ensuring this further increased level of safety internationally. These qualitative statements need to be cast into quantitative guidelines which can be used as a rationale for safety evaluation. This paper tries to describe the status reached and to stimulate international activities. (author). 12 refs, 4 figs, 3 tabs

  15. Research of the heat exchanging processes running in the heating and hot water supply loops of the coil heat exchangers

    Directory of Open Access Journals (Sweden)

    Ірина Геннадіївна Шитікова

    2016-11-01

    Full Text Available The fuel-energy complex research has made it possible to disclose a huge power-saving potential in the municipal heat-and-power engineering. Power-and-resource-saving units and systems are becoming extremely urgent because of the power engineering crisis expansion. The self-adjusting heat supply system from the individual heating points with the heat-accumulating units and coil heat exchangers for independent heating and water supply systems has been examined. Coil heat exchangers are used in municipal heating for heat transfer (e.g. geothermal waters for the independent mains of the heating and hot water supply systems. The heat engineering calculation of the heating and accumulating unit with the coil heat exchanger for independent heat supply systems from individual heater was performed and experimental data were received at the experimental industrial unit under the laboratory conditions. The peculiarities of the flows in the intertubular space, their influence on the heat exchange and temperatures of the first and intermediate mains have been shown. It is important to know the processes running inside the apparatus to be able to improve the technical characteristics of the three-loop coil heat exchanger. The task solution will make it possible to save the materials consumption for the three-loop coil heat exchangers in the future

  16. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  17. Conceptual design of small-sized HTGR system (1). Major specifications and system designs

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Yan, Xing L.; Tachibana, Yukio

    2011-06-01

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2030s. The design philosophy is that the HTR50S is a high advanced reactor, which is reducing the R and D risk based on the HTTR design, upgrading the performance and reducing the cost for commercialization by utilizing the knowledge obtained by the HTTR operation and the GTHTR300 design. The major specifications of the HTR50S were determined and targets of the technology demonstration using the HTR50S (e.g., the increasing the power density, reduction of the number of uranium enrichment in the fuel, increasing the burn up, side-by-side arrangement between the reactor pressure vessel and the steam generator) were identified. In addition, the system design of HTR50S, which offers the capability of electricity generation, cogeneration of electricity and steam for a district heating and industries, was performed. Furthermore, a market size of small-sized HTGR systems was investigated. (author)

  18. New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

    International Nuclear Information System (INIS)

    McDonald, C.F.; Sonn, D.L.

    1983-01-01

    Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR

  19. Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys

    International Nuclear Information System (INIS)

    Aronson, S.; Chow, J.G.Y.; Soo, P.; Friedlander, M.

    1978-01-01

    Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650 0 C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTGR materials to more prototypic, low concentrations of I 2 , Te 2 and CsI in the presence of low partial pressures of O 2 . These samples are scheduled for mechanical bend tests after exposure to determine the degree of embrittlement

  20. SOLTECH 92 proceedings: Solar Process Heat Program

    Energy Technology Data Exchange (ETDEWEB)

    1992-03-01

    This document is a limited Proceedings, documenting the presentations given at the symposia conducted by the US Department of Energy's (DOE) Solar Industrial Program and Solar Thermal Electrical Program at SOLTECH92. The SOLTECH92 national solar energy conference was held in Albuquerque, New Mexico during the period February 17--20, 1992. The National Renewable Energy Laboratory manages the Solar Industrial Program; Sandia National Laboratories (Albuquerque) manages the Solar Thermal Electric Program. The symposia sessions were as follows: (1) Solar Industrial Program and Solar Thermal Electric Program Overviews, (2) Solar Process Heat Applications, (3) Solar Decontamination of Water and Soil; (4) Solar Building Technologies, (5) Solar Thermal Electric Systems, (6) PV Applications and Technologies. For each presentation given in these symposia, these Proceedings provide a one- to two-page abstract and copies of the viewgraphs and/or 35mm slides utilized by the speaker. Some speakers provided additional materials in the interest of completeness. The materials presented in this document were not subjected to a peer review process.

  1. Thermal control system. [removing waste heat from industrial process spacecraft

    Science.gov (United States)

    Hewitt, D. R. (Inventor)

    1983-01-01

    The temperature of an exothermic process plant carried aboard an Earth orbiting spacecraft is regulated using a number of curved radiator panels accurately positioned in a circular arrangement to form an open receptacle. A module containing the process is insertable into the receptacle. Heat exchangers having broad exterior surfaces extending axially above the circumference of the module fit within arcuate spacings between adjacent radiator panels. Banks of variable conductance heat pipes partially embedded within and thermally coupled to the radiator panels extend across the spacings and are thermally coupled to broad exterior surfaces of the heat exchangers by flanges. Temperature sensors monitor the temperature of process fluid flowing from the module through the heat exchanges. Thermal conduction between the heat exchangers and the radiator panels is regulated by heating a control fluid within the heat pipes to vary the effective thermal length of the heat pipes in inverse proportion to changes in the temperature of the process fluid.

  2. Heat pump cycle by hydrogen-absorbing alloys to assist high-temperature gas-cooled reactor in producing hydrogen

    International Nuclear Information System (INIS)

    Satoshi, Fukada; Nobutaka, Hayashi

    2010-01-01

    A chemical heat pump system using two hydrogen-absorbing alloys is proposed to utilise heat exhausted from a high-temperature source such as a high-temperature gas-cooled reactor (HTGR), more efficiently. The heat pump system is designed to produce H 2 based on the S-I cycle more efficiently. The overall system proposed here consists of HTGR, He gas turbines, chemical heat pumps and reaction vessels corresponding to the three-step decomposition reactions comprised in the S-I process. A fundamental research is experimentally performed on heat generation in a single bed packed with a hydrogen-absorbing alloy that may work at the H 2 production temperature. The hydrogen-absorbing alloy of Zr(V 1-x Fe x ) 2 is selected as a material that has a proper plateau pressure for the heat pump system operated between the input and output temperatures of HTGR and reaction vessels of the S-I cycle. Temperature jump due to heat generated when the alloy absorbs H 2 proves that the alloy-H 2 system can heat up the exhaust gas even at 600 deg. C without any external mechanical force. (authors)

  3. Generation of a Broad-Group HTGR Library for Use with SCALE

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Lee, Deokjung; Wiarda, Dorothea; Williams, Mark L.; Mertyurek, Ugur

    2012-01-01

    With current and ongoing interest in high temperature gas reactors (HTGRs), the U.S. Nuclear Regulatory Commission (NRC) anticipates the need for nuclear data libraries appropriate for use in applications for modeling, assessing, and analyzing HTGR reactor physics and operating behavior. The objective of this work was to develop a broad-group library suitable for production analyses with SCALE for HTGR applications. Several interim libraries were generated from SCALE fine-group 238- and 999-group libraries, and the final broad-group library was created from Evaluated Nuclear Data File/B Version ENDF/B-VII Release 0 cross-section evaluations using new ORNL methodologies with AMPX, SCALE, and other codes. Furthermore, intermediate resonance (IR) methods were applied to the HTGR broadgroup library, and lambda factors and f-factors were incorporated into the library s nuclear data files. A new version of the SCALE BONAMI module named BONAMI-IR was developed to process the IR data in the new library and, thus, eliminate the need for the CENTRM/PMC modules for resonance selfshielding. This report documents the development of the HTGR broad-group nuclear data library and the results of test and benchmark calculations using the new library with SCALE. The 81-group library is shown to model HTGR cases with similar accuracy to the SCALE 238-group library but with significantly faster computational times due to the reduced number of energy groups and the use of BONAMI-IR instead of BONAMI/CENTRM/PMC for resonance self-shielding calculations.

  4. Process Heat Exchanger Options for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-06-01

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  5. Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

    International Nuclear Information System (INIS)

    Sabharwall, Piyush; Kim, Eung Soo; McKellar, Michael; Anderson, Nolan

    2011-01-01

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  6. Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-04-01

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  7. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    Energy Technology Data Exchange (ETDEWEB)

    Malevski, A L; Stoliarevski, A Ya; Vladimirov, V T; Larin, E A; Lesnykh, V V; Naumov, Yu V; Fedotov, I L

    1990-07-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

  8. The choice of equipment mix and parameters for HTGR-based nuclear cogeneration plants

    International Nuclear Information System (INIS)

    Malevski, A.L.; Stoliarevski, A.Ya.; Vladimirov, V.T.; Larin, E.A.; Lesnykh, V.V.; Naumov, Yu.V.; Fedotov, I.L.

    1990-01-01

    Improvement of heat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources conserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hydrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th) in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperature of 920-1020 K it is possible to create steam turbine installations producing both

  9. Process heat transfer principles, applications and rules of thumb

    CERN Document Server

    Serth, Robert W

    2014-01-01

    Process Heat Transfer is a reference on the design and implementation of industrial heat exchangers. It provides the background needed to understand and master the commercial software packages used by professional engineers in the design and analysis of heat exchangers. This book focuses on types of heat exchangers most widely used by industry: shell-and-tube exchangers (including condensers, reboilers and vaporizers), air-cooled heat exchangers and double-pipe (hairpin) exchangers. It provides a substantial introduction to the design of heat exchanger networks using pinch technology, the mos

  10. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  11. Overview of HTGR utilization system developments at JAERI

    International Nuclear Information System (INIS)

    Miyamoto, Y.; Shiozawa, S.; Inagaki, Y.

    1997-01-01

    JAERI has been constructing a 30-MWt HTGR, named HTTR, to develop technology and to demonstrate effectiveness of high-temperature nuclear heat utilization. A hydrogen production system by natural gas steam reforming is to be the first heat utilization system of the HTTR since its technology matured in fossil-fired plant enables to couple with HTTR in the early 2000's and technical solutions demonstrated by the coupling will contribute to all other hydrogen production systems. The HTTR steam reforming system is designed to utilize the nuclear heat effectively and to achieve hydrogen productivity competitive to that of a fossil-fired plant with operability, controllability and safety acceptable enough to commercialization, and an arrangement of key components was already decided. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test is planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions. The out-of-pile system is an approximately 1/20-1/30 scale system of the HTTR steam reforming system and simulates key components downstream from an IHX

  12. Availability of steam generator against thermal disturbance of hydrogen production system coupled to HTGR

    International Nuclear Information System (INIS)

    Shibata, Taiju; Nishihara, Tetsuo; Hada, Kazuhiko; Shiozawa, Shusaku

    1996-01-01

    One of the safety issues to couple a hydrogen production system to an HTGR is how the reactor coolability can be maintained against anticipated abnormal reduction of heat removal (thermal disturbance) of the hydrogen production system. Since such a thermal disturbance is thought to frequently occur, it is desired against the thermal disturbance to keep reactor coolability by means other than reactor scram. Also, it is thought that the development of a passive cooling system for such a thermal disturbance will be necessary from a public acceptance point of view in a future HTGR-hydrogen production system. We propose a SG as the passive cooling system which can keep the reactor coolability during a thermal disturbance of a hydrogen production system. This paper describes the proposed steam generator (SG) for the HTGR-hydrogen production system and a result of transient thermal-hydraulic analysis of the total system, showing availability of the SG against a thermal disturbance of the hydrogen production system in case of the HTTR-steam reforming hydrogen production system. (author)

  13. Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y., E-mail: fukaya.yuji@jaea.go.jp; Goto, M.; Nishihara, T.

    2015-11-15

    Highlights: • We attempted and optimized erbium loading methods to improve reactivity coefficients for LRSF-HTGR. • We elucidated the mechanism of the improvements for each erbium loading method by using the Bondarenko approach. • We concluded the erbium loading method by embedding into graphite shaft is preferable. - Abstract: Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact. It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach. As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

  14. Consideration on developing of leaked inflammable gas detection system for HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Nakamura, Masashi

    1999-09-01

    One of most important safety design issues for High Temperature Gas-cooled Reactor (HTGR) - Hydrogen Production System (HTGR-HPS) is to ensure reactor safety against fire and explosion at the hydrogen production plant. The inflammable gas mixture in the HTGR-HPS does not use oxygen in any condition and are kept in high pressure in the normal operation. The piping system and/or heat transfer tubes which have the potential possibility of combustible materials ingress into the Reactor Building (R/B) due to the failure are designed to prevent the failure against any events. Then, it is not necessary to consider their self-combustion in vessels nor leakage in the R/B. The only one case which we must consider is the ex-building fire or explosion caused by their leakage from piping or vessel. And it is important to mitigate their effects by means of early detection of gas leakage. We investigated our domestic standards on gas detection, applications of gas detectors, their detection principles, performance, sensitivity, reliability, their technical trends, and so on. We proposed three gas detection systems which may be applied in HTGR-HPS. The first one is the universal solid sensor system; it may be applied when there is no necessity to request their safety credits. The second is the combination of the improved solid sensor system and enhanced beam detector system; it may be applied when it is necessary to request their safety credit. And the third is the combination of the universal solid sensor system and the existing beam detector system; it may be applied when the plant owner request higher detector sensitivity than usual, from the view point of public acceptance, though there is not necessity to request their safety credits. To reduce the plant cost by refusing of safety credits to the gas leakage detection system, we proposed that the equipment required to isolate from others should be installed in the inertrized compartments. (author)

  15. HTR process heat applications, status of technology and economical potential

    International Nuclear Information System (INIS)

    Barnet, H.

    1997-01-01

    The technical and industrial feasibility of the production of high temperature heat from nuclear fuel is presented. The technical feasibility of high temperature heat consuming processes is reviewed and assessed. The conclusion is drawn that the next technological step for pilot plant scale demonstration is the nuclear heated steam reforming process. The economical potential of HTR process heat applications is reviewed: It is directly coupled to the economical competitiveness of HTR electricity production. Recently made statements and pre-conditions on the economic competitiveness in comparison to world market coal are reported. (author). 8 figs

  16. HTGR Metallic Reactor Internals Core Shell Cutting & Machining Antideformation Technique Study

    International Nuclear Information System (INIS)

    Xing Huiping; Xue Song

    2014-01-01

    The reactor shell assembly of HTGR nuclear power station demonstration project metallic reactor internals is key components of reactor, remains with high-precision large component with large-sized thin-walled straight cylinder-shaped structure, and is the first manufacture in China. As compared with other reactor shell, it has a larger ID (Φ5360mm), a longer length (19000mm), a smaller wall thickness (40mm) and a higher precision requirement. During the process of manufacture, the deformation due to cutting & machining will directly affect the final result of manufacture, the control of structural deformation and cutting deformation shall be throughout total manufacture process of such assembly. To realize the control of entire core shell assembly geometry, the key is to innovate and make breakthroughs on anti-deformation technique and then provide reliable technological foundations for the manufacture of HTGR metallic reactor internals. (author)

  17. Dynamic response of a multielement HTGR core

    International Nuclear Information System (INIS)

    Reich, M.; Bezler, P.; Koplik, B.; Curreri, J.; Goradia, H.; Lasker, L.

    1977-01-01

    One of the primary factors in determining the structural integrity and consequently the safety of a High Temperature Gas-Cooled Reactor (HTGR) is the dynamic response of the core when subjected to a seismic excitation. The HTGR core under consideration consists of several thousands of hexagonal elements arranged in vertical stacks containing about eight elements per stack. There are clearance gaps between adjacent elements, which can change substantially due to radiation effects produced during their active lifetime. Surrounding the outer periphery of the core are reflector blocks and restraining spring-pack arrangements which bear against the reactor vessel structure (PCRV). Earthquake input motions to this type of core arrangement will result in multiple impacts between adjacent elements as well as between the reflector blocks and the restraining spring packs. The highly complex nonlinear response associated with the multiple collisions across the clearance gaps and with the spring packs is the subject matter of this paper. Of particular importance is the ability to analyze a complex nonlinear system with gaps by employing a model with a reduced number of masses. This is necessary in order to obtain solutions in a time-frame and at a cost which is not too expensive. In addition the effect of variations in total clearance as well as the initial distribution of clearances between adjacent elements is of primary concern. Both of these aspects of the problem are treated in the present analysis. Finally, by constraining the motion of the reflector blocks, a more realistic description of the dynamic response of the multi-element HTGR core is obtained

  18. Heat exchanger for coal gasification process

    Science.gov (United States)

    Blasiole, George A.

    1984-06-19

    This invention provides a heat exchanger, particularly useful for systems requiring cooling of hot particulate solids, such as the separated fines from the product gas of a carbonaceous material gasification system. The invention allows effective cooling of a hot particulate in a particle stream (made up of hot particulate and a gas), using gravity as the motive source of the hot particulate. In a preferred form, the invention substitutes a tube structure for the single wall tube of a heat exchanger. The tube structure comprises a tube with a core disposed within, forming a cavity between the tube and the core, and vanes in the cavity which form a flow path through which the hot particulate falls. The outside of the tube is in contact with the cooling fluid of the heat exchanger.

  19. Dependable Hydrogen and Industrial Heat Generation from the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Charles V. Park; Michael W. Patterson; Vincent C. Maio; Piyush Sabharwall

    2009-03-01

    The Department of Energy is working with industry to develop a next generation, high-temperature gas-cooled nuclear reactor (HTGR) as a part of the effort to supply the US with abundant, clean and secure energy. The Next Generation Nuclear Plant (NGNP) project, led by the Idaho National Laboratory, will demonstrate the ability of the HTGR to generate hydrogen, electricity, and high-quality process heat for a wide range of industrial applications. Substituting HTGR power for traditional fossil fuel resources reduces the cost and supply vulnerability of natural gas and oil, and reduces or eliminates greenhouse gas emissions. As authorized by the Energy Policy Act of 2005, industry leaders are developing designs for the construction of a commercial prototype producing up to 600 MWt of power by 2021. This paper describes a variety of critical applications that are appropriate for the HTGR with an emphasis placed on applications requiring a clean and reliable source of hydrogen. An overview of the NGNP project status and its significant technology development efforts are also presented.

  20. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power

  1. Quantitative HTGR safety and forced outage goals

    International Nuclear Information System (INIS)

    Houghton, W.J.; Parme, L.L.; Silady, F.A.

    1985-05-01

    A key step in the successful implementation of the integrated approach is the definition of the overall plant-level goals. To be effective, the goals should provide clear statements of what is to be achieved by the plant. This can be contrasted to the current practice of providing design-prescriptive criteria which implicitly address some higher-level objective but restrict the designer's flexibility. Furthermore, the goals should be quantifiable in such a way that satisfaction of the goal can be measured. In the discussion presented, two such plant-level goals adopted for the HTGR and addressing the impact of unscheduled occurrences are described. 1 fig

  2. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power.

  3. Heat flow and geothermal processes in Iceland

    Science.gov (United States)

    Flóvenz, Ólafur G.; Saemundsson, Kristján

    1993-09-01

    Heat flow values, derived from temperature measurements in shallow boreholes in Iceland, vary substantially across the country. The near-surface temperature gradients range from almost 0 to 500°C/km. The thermal conductivity of water-saturated rocks varies from 1.6 to 2.0 W/m°C. The temperature gradient in Iceland is mainly dependent on four factors: (1) the regional heat flow through the crust, (2) hydrothermal activity, (3) the permeability of the rock, and (4) residual heat in extinct volcanic centers. As Iceland is mainly made of basaltic material the radiogenic heat production is almost negligible. The thermal conductivity is, on the other hand, mainly influenced by the porosity of the rock; it increases as the porosity decreases. Iceland is made of sequences of flood basalts that formed within the volcanic rift zone—a continuation of the axis of the Mid-Atlantic ridge—and subsequently drifted sideways. Fresh basaltic lava is usually highly porous (30%) and fractured, and heat is mainly transported by convection. Therefore, a very low or even no temperature gradient is observed at shallow levels within the volcanic rift zone. As the basalt becomes buried the pores close due to lithostatic pressure and formation of secondary minerals. Below 500-1000 m depth in an uneroded lava pile, the heat is mainly transported by conduction. In the lowlands and valleys of Iceland outside the volcanic rift zone, 1000-1500 m of the original lava pile has been eroded, leaving thermal conduction as the most important heat transport mechanism. The regional temperature gradient has been measured in drillholes in dense and poorly permeable rocks away from the geothermal fields. The results show that the temperature gradient varies from 50 to 150°C/km. The highest values are found close to the volcanic rift zone and the gradient decreases with distance from the spreading axis. This result is mainly based on numerous shallow boreholes (60-500 m) but in some cases the results

  4. Process heat cogeneration using a high temperature reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  6. Gasification of coal making use of nuclear processing heat

    International Nuclear Information System (INIS)

    Schilling, H.D.; Bonn, B.; Krauss, U.

    1981-01-01

    In the chapter 'Gasification of coal making use of nuclear processing heat', the steam gasification of brown coal and bituminous coal, the hydrogenating gasification of brown coal including nuclear process heat either by steam cracking methane in the steam reformer or by preheating the gasifying agent, as well as the hydrogenating gasification of bituminous coal are described. (HS) [de

  7. Process for adapting a heat source and a thermal machine by temporary heat storage

    International Nuclear Information System (INIS)

    Cahn, R.P.; Nicholson, E.W.

    1975-01-01

    The process described is intended to ensure the efficient use of the heat from a nuclear reactor or from a furnace burning fossil fuel at constant power, and of a boiler in a power station comprising a multi-stage steam turbine, the steam extracted from the turbine being used for pre-heating the boiler feed water. This process is most flexible with a varying load. It includes the high temperature storage of the excess heat energy in a low vapor pressure storage liquid (hydrocarbon oils, molten salts or liquid metals) at atmospheric pressure when the demand is low; then, when the energy demand is at its height, the reduction of steam extraction from the turbine with simultaneous utilisation of the hot heat storage liquid for the various maintenance heating functions of the power station by heat exchange, so that the heat can expand totally in the turbine with generation of energy [fr

  8. Role of the HTGR in the U.S. industrial energy market

    International Nuclear Information System (INIS)

    Leeth, G.G.

    1981-01-01

    The HTGR is considered for a variety of applications to the U.S. industrial energy markets. These include a number of synfuel processes, shale oil conversion, methanol production, ammonia production, and both open and closed-loop pipeline systems. Potential market size appears to be approximately 300-400 GW (t) in the 2000 to 2020 time period. In addition to potential cost advantages, the closed-loop nuclear system has several significant advantages over alternative fossil systems. 5 refs

  9. Process for heating coal-oil slurries

    Science.gov (United States)

    Braunlin, W.A.; Gorski, A.; Jaehnig, L.J.; Moskal, C.J.; Naylor, J.D.; Parimi, K.; Ward, J.V.

    1984-01-03

    Controlling gas to slurry volume ratio to achieve a gas holdup of about 0.4 when heating a flowing coal-oil slurry and a hydrogen containing gas stream allows operation with virtually any coal to solvent ratio and permits operation with efficient heat transfer and satisfactory pressure drops. The critical minimum gas flow rate for any given coal-oil slurry will depend on numerous factors such as coal concentration, coal particle size distribution, composition of the solvent (including recycle slurries), and type of coal. Further system efficiency can be achieved by operating with multiple heating zones to provide a high heat flux when the apparent viscosity of the gas saturated slurry is highest. Operation with gas flow rates below the critical minimum results in system instability indicated by temperature excursions in the fluid and at the tube wall, by a rapid increase and then decrease in overall pressure drop with decreasing gas flow rate, and by increased temperature differences between the temperature of the bulk fluid and the tube wall. At the temperatures and pressures used in coal liquefaction preheaters the coal-oil slurry and hydrogen containing gas stream behaves essentially as a Newtonian fluid at shear rates in excess of 150 sec[sup [minus]1]. The gas to slurry volume ratio should also be controlled to assure that the flow regime does not shift from homogeneous flow to non-homogeneous flow. Stable operations have been observed with a maximum gas holdup as high as 0.72. 29 figs.

  10. Frictional heating processes during laboratory earthquakes

    Science.gov (United States)

    Aubry, J.; Passelegue, F. X.; Deldicque, D.; Lahfid, A.; Girault, F.; Pinquier, Y.; Escartin, J.; Schubnel, A.

    2017-12-01

    Frictional heating during seismic slip plays a crucial role in the dynamic of earthquakes because it controls fault weakening. This study proposes (i) to image frictional heating combining an in-situ carbon thermometer and Raman microspectrometric mapping, (ii) to combine these observations with fault surface roughness and heat production, (iii) to estimate the mechanical energy dissipated during laboratory earthquakes. Laboratory earthquakes were performed in a triaxial oil loading press, at 45, 90 and 180 MPa of confining pressure by using saw-cut samples of Westerly granite. Initial topography of the fault surface was +/- 30 microns. We use a carbon layer as a local temperature tracer on the fault plane and a type K thermocouple to measure temperature approximately 6mm away from the fault surface. The thermocouple measures the bulk temperature of the fault plane while the in-situ carbon thermometer images the temperature production heterogeneity at the micro-scale. Raman microspectrometry on amorphous carbon patch allowed mapping the temperature heterogeneities on the fault surface after sliding overlaid over a few micrometers to the final fault roughness. The maximum temperature achieved during laboratory earthquakes remains high for all experiments but generally increases with the confining pressure. In addition, the melted surface of fault during seismic slip increases drastically with confining pressure. While melting is systematically observed, the strength drop increases with confining pressure. These results suggest that the dynamic friction coefficient is a function of the area of the fault melted during stick-slip. Using the thermocouple, we inverted the heat dissipated during each event. We show that for rough faults under low confining pressure, less than 20% of the total mechanical work is dissipated into heat. The ratio of frictional heating vs. total mechanical work decreases with cumulated slip (i.e. number of events), and decreases with

  11. Industrial process heat usage in the United States

    International Nuclear Information System (INIS)

    1981-03-01

    The purpose of this report is to identify and evaluate sources of information on industrial energy consumption, which can serve as a reference for assessing the market potential of HTGR programs and thereby provide a consistent information base for all applications programs. The report provides information on interpretation of energy use data, definition of the industrial sector, energy use by the industrial sector and other sectors, and use within the industrial sector by industry type and service demand. The report also reviews several data sources and presents historical data and projections on industrial energy consumption. These data and projections are taken from the sources which appear to be most representative of the actual market and most useful with respect to manipulation of data to provide information needed for HTGR programs

  12. Present status of HTGR research and development

    International Nuclear Information System (INIS)

    1991-04-01

    The HTTR is a test reactor with thermal output of 30MW and outlet coolant temperature of 950degC, employing the pin-in-block type fuel, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. The official construction of the HTTR facility is scheduled to start on March 15, 1991. This publication summarizes the present status of R and D of high temperature gas cooled reactors in JAERI. (J.P.N.)

  13. User's manual for the Composite HTGR Analysis Program (CHAP-1)

    International Nuclear Information System (INIS)

    Gilbert, J.S.; Secker, P.A. Jr.; Vigil, J.C.; Wecksung, M.J.; Willcutt, G.J.E. Jr.

    1977-03-01

    CHAP-1 is the first release version of an HTGR overall plant simulation program with both steady-state and transient solution capabilities. It consists of a model-independent systems analysis program and a collection of linked modules, each representing one or more components of the HTGR plant. Detailed instructions on the operation of the code and detailed descriptions of the HTGR model are provided. Information is also provided to allow the user to easily incorporate additional component modules, to modify or replace existing modules, or to incorporate a completely new simulation model into the CHAP systems analysis framework

  14. Beginning-of-life neutronic analysis of a 3000-MW(t) HTGR

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1975-12-01

    The results of a study of safety-related neutronic characteristics for the beginning-of-life core of a 3000-MW(t) High-Temperature Gas-Cooled Reactor are presented. Emphasis was placed on the temperature-dependent reactivity effects of fuel, moderator, control poisons, and fission products. Other neutronic characteristics studied were gross and local power distributions, neutron kinetics parameters, control rod and other material worths and worth distributions, and the reactivity worth of a selected hypothetical perturbation in the core configuration. The study was performed for the most part using discrete-ordinates transport theory codes and neutron cross sections that were interpolated from a four-parameter nine-group library supplied by the HTGR vendor. A few comparison calculations were also performed using nine-group data generated with an independent cross-section processing code system. Results from the study generally agree well with results reported by the HTGR vendor

  15. SONATINA-1: a computer program for seismic response analysis of column in HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1980-11-01

    An computer program SONATINA-1 for predicting the behavior of a prismatic high-temperature gas-cooled reactor (HTGR) core under seismic excitation has been developed. In this analytical method, blocks are treated as rigid bodies and are constrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions. Coulomb friction between blocks and between dowel holes and pins is also considered. A spring dashpot model is used for the collision process between adjacent blocks and between blocks and boundary walls. Analytical results are compared with experimental results and are found to be in good agreement. The computer program can be used to predict the behavior of the HTGR core under seismic excitation. (author)

  16. A hybrid HTGR system producing electricity, hydrogen and such other products as water demanded in the Middle East

    Energy Technology Data Exchange (ETDEWEB)

    Yan, X., E-mail: yan.xing@jaea.go.jp; Noguchi, H.; Sato, H.; Tachibana, Y.; Kunitomi, K.; Hino, R.

    2014-05-01

    Alternative energy products are being considered by the Middle East countries for both consumption and export. Electricity, water, and hydrogen produced not from oil and gas are amongst those desirable. A hybrid nuclear production system, GTHTR300C, under development in JAEA can achieve this regional strategic goal. The system is based on a 600 MWt HTGR and equipped to cogenerate electricity by gas turbine and seawater desalination by using only the nuclear plant waste heat. Hydrogen is produced via a thermochemical water-splitting process driven by the reactor's 950 °C heat. Additionally process steam may be produced for industrial uses. An example is shown of manufacturing soda ash, an internationally traded commodity, from using the steam produced and the brine discharged from desalination. The nuclear reactor satisfies nearly all energy requirements for the hybrid generations without emitting CO{sub 2}. The passive safety of the reactor as described in the paper permits proximity of siting the reactor with the production facilities to enhance energy transmission. Production flowsheet of the GTHTR300C is given for up to 300 MWe electricity, 58 t/day hydrogen, 56,000 m{sup 3}/day potable water, 3500 t/day steam, and 1000 t/day soda ash. The production thermal efficiency reaches 88%.

  17. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  18. Energy analysis of a desalination process of sea water with nuclear energy

    International Nuclear Information System (INIS)

    Martinez L, G.; Valle H, J.

    2016-09-01

    In the present work, is theoretically proven that the residual heat, removed by the chillers in the stage prior to the compression of the recuperative Brayton cycle with which nuclear power plants operate with high temperature gas reactors (HTGR), can be used to produce stem and desalinate seawater. The desalination process selected for the analysis, based on its operating characteristics, is the Multi-Stage Distillation (Med). The Med process will use as energy source, for the flash evaporation process in the flash trap, the residual heat that the reactor coolant dissipates to the environment in order to increase the compression efficiency of the same; the energy dissipated depends on the operating conditions of the reactor. The Med distillation process requires saturated steam at low pressure which can be obtained by means of a heat exchanger, taking advantage of the residual heat, where the relative low temperatures with which the process operates make the nuclear plants with HTGR reactors ideal for desalination of sea water, because they do not require major modifications to their design of their operation. In this work the energy analysis of a six-stage Med module coupled to the chillers of an HTGR reactor of the Pebble Bed Modular Reactor type is presented. Mathematical modeling was obtained by differential equations of mass and energy balances in the system. The results of the analysis are presented in a table for each distillation stage, estimating the pure water obtained as a function of the heat supplied. (Author)

  19. Evaluating the potential of process sites for waste heat recovery

    International Nuclear Information System (INIS)

    Oluleye, Gbemi; Jobson, Megan; Smith, Robin; Perry, Simon J.

    2016-01-01

    Highlights: • Analysis considers the temperature and duties of the available waste heat. • Models for organic Rankine cycles, absorption heat pumps and chillers proposed. • Exploitation of waste heat from site processes and utility systems. • Concept of a site energy efficiency introduced. • Case study presented to illustrate application of the proposed methodology. - Abstract: As a result of depleting reserves of fossil fuels, conventional energy sources are becoming less available. In spite of this, energy is still being wasted, especially in the form of heat. The energy efficiency of process sites (defined as useful energy output per unit of energy input) may be increased through waste heat utilisation, thereby resulting in primary energy savings. In this work, waste heat is defined and a methodology developed to identify the potential for waste heat recovery in process sites; considering the temperature and quantity of waste heat sources from the site processes and the site utility system (including fired heaters and, the cogeneration, cooling and refrigeration systems). The concept of the energy efficiency of a site is introduced – the fraction of the energy inputs that is converted into useful energy (heat or power or cooling) to support the methodology. Furthermore, simplified mathematical models of waste heat recovery technologies using heat as primary energy source, including organic Rankine cycles (using both pure and mixed organics as working fluids), absorption chillers and absorption heat pumps are developed to support the methodology. These models are applied to assess the potential for recovery of useful energy from waste heat. The methodology is illustrated for an existing process site using a case study of a petroleum refinery. The energy efficiency of the site increases by 10% as a result of waste heat recovery. If there is an infinite demand for recovered energy (i.e. all the recoverable waste heat sources are exploited), the site

  20. Process for preparing a normal lighting and heating gas etc

    Energy Technology Data Exchange (ETDEWEB)

    Becker, J

    1910-12-11

    A process for preparing a normal lighting and heating gas from Australian bituminous shale by distillation and decomposition in the presence of water vapor is characterized by the fact that the gasification is suitably undertaken with gradual filling of a retort and with simultaneous introduction of water vapor at a temperature not exceeding 1,000/sup 0/ C. The resulting amount of gas is heated in the same or a second heated retort with freshly supplied vapor.

  1. Process for forming thin film, heat treatment process of thin film sheet, and heat treatment apparatus therefor

    International Nuclear Information System (INIS)

    Watanabe, S.

    1984-01-01

    The invention provides a process for forming a magnetic thin film on a base film, a heat treatment process of a thin film sheet consisting of the base film and the magnetic thin film, and an apparatus for performing heat treatment of the thin film sheet. Tension applied to the thin film sheet is substantially equal to that applied to the base film when the magnetic thin film is formed thereon. Then, the thin film sheet is treated with heat. The thin film sheet is heated with a given temperature gradient to a reactive temperature at which heat shrinkage occurs, while the tension is being applied thereto. Thereafter, the thin film sheet to which the tension is still applied is cooled with substantially the same temperature gradient as applied in heating. The heat treatment apparatus has a film driving unit including a supply reel, a take-up reel, a drive source and guide rollers; a heating unit including heating plates, heater blocks and a temperature controller for heating the sheet to the reactive temperature; and a heat insulating unit including a thermostat and another temperature controller for maintaining the sheet at the nonreactive temperature which is slightly lower than the reactive temperature

  2. Analytical models of Ohmic heating and conventional heating in food processing

    Science.gov (United States)

    Serventi, A.; Bozzoli, F.; Rainieri, S.

    2017-11-01

    Ohmic heating is a food processing operation in which an electric current is passed through a food and the electrical resistance of the food causes the electric power to be transformed directly into heat. The heat is not delivered through a surface as in conventional heat exchangers but it is internally generated by Joule effect. Therefore, no temperature gradient is required and it origins quicker and more uniform heating within the food. On the other hand, it is associated with high energy costs and its use is limited to a particular range of food products with an appropriate electrical conductivity. Sterilization of foods by Ohmic heating has gained growing interest in the last few years. The aim of this study is to evaluate the benefits of Ohmic heating with respect to conventional heat exchangers under uniform wall temperature, a condition that is often present in industrial plants. This comparison is carried out by means of analytical models. The two different heating conditions are simulated under typical circumstances for the food industry. Particular attention is paid to the uniformity of the heat treatment and to the heating section length required in the two different conditions.

  3. Mini-channel heat exchangers for industrial distillation processes

    NARCIS (Netherlands)

    Van de Bor, D.M.

    2014-01-01

    In this thesis the technical and economic performance of compression-resorption heat pumps has been investigated. The main objective of this thesis was to improve the performance and reduce the investment costs of compression-resorption heat pumps applied in process industry. A model that is able to

  4. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

    Reprocessing of HTGR fuels requires evaporative concentration of uranium and thorium nitrate solutions. The results of a bench-scale test program conducted to assess the safety aspects of planned concentrator operations are reported

  5. HTGR safety research program. Progress report, April--June 1975

    International Nuclear Information System (INIS)

    Kirk, W.L.

    1975-09-01

    Progress in HTGR safety research is reported under the following headings: fission product technology; primary coolant impurities; structural investigation; safety instrumentation and control systems; phenomena modeling and systems analysis. (JWR)

  6. Status of the United States National HTGR program

    International Nuclear Information System (INIS)

    1981-01-01

    The HTGR continues to appear as an increasingly attractive option for application to US energy markets. To examine that potential, a program is being pursued to examine the various HTGR applications and to provide information to decision-makers in both the public and private sectors. To date, this effort has identified a substantial technical and economic potential for Steam Cycle/Cogeneration applications. Advanced HTGR systems are currently being evaluated to determine their appropriate role and timing. The encouraging results which have been obtained lead to heightened anticipation that a role for the HTGR will be found in the US energy market and that an initiative culminating in a lead project will be evolved in the forseeable future. The US Program can continue to benefit from international cooperative activities to develop the needed technologies. Expansion of these cooperative activities will be actively pursued

  7. Effects of graphite surface roughness on bypass flow computations for an HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Yu-Hsin, E-mail: touushin@gmail.com [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Johnson, Richard W., E-mail: Rich.Johnson@inl.gov [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Sato, Hiroyuki, E-mail: sato.hiroyuki09@jaea.go.jp [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer CFD calculations are made of bypass flow between graphite blocks in HTGR. Black-Right-Pointing-Pointer Several turbulence models are employed to compare to friction and heat transfer correlations. Black-Right-Pointing-Pointer Parameters varied include bypass gap width and surface roughness. Black-Right-Pointing-Pointer Surface roughness causes increases in max fuel and coolant temperatures. Black-Right-Pointing-Pointer Surface roughness does not cause increase in outlet coolant temperature variation. - Abstract: Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow; it has been suggested that it may be as much as 20% of the total helium coolant flow [INL Report 2007, INL/EXT-07-13289]. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors for three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U.S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for steady flow in a

  8. Influence of microwave heating on the stability of processed samn

    OpenAIRE

    Farag, Radwan S.; Taha, Soad H.

    1991-01-01

    Butter was converted to samn by microwave and conventional heating. The quality of the processed samn by the two methods was followed by determining the acid, peroxide and TBA values over a period of six weeks at 60°C. The fatty acid composition of samn samples was determined by gas-liquid chromatographic technique. The data show that butter conversion to samn by microwave heating was accomplished in about one half of the time that conventional heating requires. Microwave heating obviously in...

  9. Conceptual design of primary coolant purification system using cylindrical membrane for nuclear energy system base on HTGR

    International Nuclear Information System (INIS)

    Piping Supriatna

    2011-01-01

    The recent progress of reactor technology design for next generation reactor will be implemented on cogeneration reactor, which the aim of reactor operation not only for generating electrical energy, but also for other application like desalination, industrial manufacturing process, hydrogen production, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor concept developed for generate energy effectively, efficiently and sustainable, which reserve of uranium and thorium nuclear fuel for cogeneration reactor is supply able for world energy demand until next thousand years. The cogeneration reactor produce temperature output higher than commonly Nuclear Power Plant (NPP), and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this research has been designed modeling and assessment of primary coolant gas purification system with purify and fill up helium gas continuously, by using Cylindrical Helium Splitting Membrane and helium gas inventory system. The result of flow rate helium assessment for the purification system is 0.844x10 -3 kg/sec, where helium flow rate of reactor primary coolant is 120 kg/sec. The result of study show that the Primary Coolant Gas Purification System is enable to be implemented on Cogeneration Reactor HTGR200C. (author)

  10. Volume 2. Probabilistic analysis of HTGR application studies. Supporting data

    International Nuclear Information System (INIS)

    1980-09-01

    Volume II, Probabilistic Analysis of HTGR Application Studies - Supporting Data, gives the detail data, both deterministic and probabilistic, employed in the calculation presented in Volume I. The HTGR plants and the fossil plants considered in the study are listed. GCRA provided the technical experts from which the data were obtained by MAC personnel. The names of the technical experts (interviewee) and the analysts (interviewer) are given for the probabilistic data

  11. Characteristics of radioactive waste streams generated in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Lin, K.H.

    1976-01-01

    Results are presented of a study concerned with identification and characterization of radioactive waste streams from an HTGR fuel reprocessing plant. Approximate quantities of individual waste streams as well as pertinent characteristics of selected streams have been estimated. Most of the waste streams are unique to HTGR fuel reprocessing. However, waste streams from the solvent extraction system and from the plant facilities do not differ greatly from the corresponding LWR fuel reprocessing wastes

  12. 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)

  13. Recent trends and developments in infrared heating in food processing.

    Science.gov (United States)

    Rastogi, Navin K

    2012-01-01

    Fruit processing and preservation technologies must keep fresh-like characteristics while providing an acceptable and convenient shelf life as well as assuring safety and nutritional value. Processing technologies include a wide range of methodologies to inactivate microorganisms, improve quality and stability, and preserve and minimize changes of fruit fresh-like characteristics. Infrared (IR) heating offers many advantages over conventional heating under similar conditions, which include reduced heating time, uniform heating, reduced quality losses, versatile, simple and compact equipment, and significant energy saving. The integration of IR with other matured processing operations such as blanching, dehydration, freeze-dehydration, thawing, roasting, baking, cooking has been shown to open up new processing options. Combinations of IR heating with microwave heating and other common conductive and convective modes of heating have been gaining momentum because of increased energy throughput. A number of publications and patents have demonstrated novel and diverse uses of this technology. This review aims at identifying the opportunities and challenges associated with this technology. The effect of IR on food quality attributes is also discussed. The types of equipment commonly used for IR processing have also been summarized.

  14. Effect of heat processing on the proximate composition and energy ...

    African Journals Online (AJOL)

    Dr J. T. Ekanem

    Received 5 August 2006. MS/No BKM/2006/027, ... In each of these locations, heat processing generally increased moisture ... underground water with hydrocarbons and dispersant products1. ..... Technology of Yam Tubers, Vol. 1. ed by.

  15. National need for utilizing nuclear energy for process heat generation

    International Nuclear Information System (INIS)

    Gambill, W.R.; Kasten, P.R.

    1984-01-01

    Nuclear reactors are potential sources for generating process heat, and their applications for such use economically competitive. They help satisfy national needs by helping conserve and extend oil and natural gas resources, thus reducing energy imports and easing future international energy concerns. Several reactor types can be utilized for generating nuclear process heat; those considered here are light water reactors (LWRs), heavy water reactors (HWRs), gas-cooled reactors (GCRs), and liquid metal reactors (LMRs). LWRs and HWRs can generate process heat up to 280 0 C, LMRs up to 540 0 C, and GCRs up to 950 0 C. Based on the studies considered here, the estimated process heat markets and the associated energy markets which would be supplied by the various reactor types are summarized

  16. Oxidation parameters of nuclear graphite for HTGR air-ingress

    International Nuclear Information System (INIS)

    Kim, E.S.; No, H.C.

    2004-01-01

    In order to investigate chemical behaviors of the graphite during an air-ingress accident in HTGR, the kinetic tests on nuclear graphite IG-110 were performed in chemical reaction dominant regime. In the present experiment, inlet gas flow rate ranged between 8 and 18 SLPM, graphite temperatures and oxygen mole fraction ranged from 540 to 630degC and from 3 to 30% respectively. The test section was made of a quartz tube having 75 mm diameter and 750 mm length and the test specimen machined to the size of 21 mm diameter and 30 mm length was supported at the center of it by the alumina rod. The 15 kW induction heater was installed around the outside of test section to heat the specimen and its temperature was measured by 2 infrared thermometers. The oxidation rate was calculated from the gas concentration analysis between inlet and outlet using NDIR (non-dispersive infrared) gas analyzer. As a result the activation energy (Ea) and the order of reaction (n) were determined within 95% confidence level and the qualitative characteristics of the two parameters were also widely investigated by experimental and analytical methods. (author)

  17. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  18. Fission-product SiC reaction in HTGR fuel

    International Nuclear Information System (INIS)

    Montgomery, F.

    1981-01-01

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

  19. Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

    2016-10-15

    Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

  20. Proceedings of the solar industrial process heat symposium

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

    The purpose of the symposium was to review the progress of various solar energy systems currently under design for supplying industrial process heat. Formal presentations consisted of a review of solar energy applications in industrial process heat as well as several on-going project reviews. An Open Forum was held to solicit the comments of the participants. The recommendations of this Open Forum are included in these proceedings. Eighteen papers were included. Separate abstracts were prepared for each paper.

  1. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

    Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

    1990-01-01

    Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

  2. The acoustic environment in large HTGR's

    International Nuclear Information System (INIS)

    Burton, T.E.

    1979-01-01

    Well-known techniques for estimating acoustic vibration of structures have been applied to a General Atomic high-temperature gas-cooled reactor (HTGR) design. It is shown that one must evaluate internal loss factors for both fluid and structure modes, as well as radiation loss factors, to avoid large errors in estimated structural response. At any frequency above 1350 rad/s there are generally at least 20 acoustic modes contributing to acoustic pressure, so statistical energy analysis may be employed. But because the gas circuit consists mainly of high-aspect-ratio cavities, reverberant fields are nowhere isotropic below 7500 rad/s, and in some regions are not isotropic below 60 000 rad/s. In comparison with isotropic reverberant fields, these anistropic fields enhance the radiation efficiencies of some structural modes at low frequencies, but have surprisingly little effect at most frequencies. The efficiency of a dipole sound source depends upon its orientation. (Auth.)

  3. HTGR strategy for reduced proliferation potential

    International Nuclear Information System (INIS)

    Stewart, H.B.; Dahlberg, R.C.

    1978-01-01

    The HTGR stratregy for reduced proliferation potential is one aspect of a potential broader nuclear strategy aimed primarily toward a transition nuclear period between today's uranium-consumption reactors and the long-range balanced system of breeder and advanced near-breeder reactors. In particular, the normal commerce of U-233 could be made acceptable by: (a) dependence on the gamma radiation from U-232 daughter products, (b) enhancement of that radioactivity by incomplete fission-product decontamination of the bred-fuel, or (c) denaturing of the U-233 with U-238. These approaches would, of course, supplement institutional initiatives to improve proliferation resistance such as the collocation of facilities and the establishment of secure energy centers. 6 refs

  4. Calorimetric assay of HTGR fuel samples

    International Nuclear Information System (INIS)

    Allen, E.J.; McNeany, S.R.; Jenkins, J.D.

    1979-04-01

    A calorimeter using a neutron source was designed and fabricated by Mound Laboratory, according to ORNL specifications. A calibration curve of the device for HTGR standard fuel rods was experimentally determined. The precision of a single measurement at the 95% confidence level was estimated to be +-0.8 μW. For a fuel sample containing 0.3 g 235 U and a neutron source containing 691 μg 252 Cf, this represents a relative standard deviation of 0.5%. Measurement time was approximately 5.5 h per sample. Use of the calorimeter is limited by its relatively poor precision, long measurement time, manual sample changing, sensitivity to room environment, and possibility of accumulated dust blocking water flow through the calorimeter. The calorimeter could be redesigned to resolve most of these difficulties, but not without significant development work

  5. HTGR-GT systems optimization studies

    International Nuclear Information System (INIS)

    Kammerzell, L.L.; Read, J.W.

    1980-06-01

    The compatibility of the inherent features of the high-temperature gas-cooled reactor (HTGR) and the closed-cycle gas turbine combined into a power conversion system results in a plant with characteristics consistent with projected utility needs and national energy goals. These characteristics are: (1) plant siting flexibility; (2) high resource utilization; (3) low safety risks; (4) proliferation resistance; and (5) low occupational exposure for operating and maintenance personnel. System design and evaluation studies on dry-cooled intercooled and nonintercooled commercial plants in the 800-MW(e) to 1200-MW(e) size range are described, with emphasis on the sensitivity of plant design objectives to variation of component and plant design parameters. The impact of these parameters on fuel cycle, fission product release, total plant economics, sensitivity to escalation rates, and plant capacity factors is examined

  6. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  7. Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

    International Nuclear Information System (INIS)

    Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

    2014-01-01

    Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

  8. Peach Bottom HTGR decommissioning and component removal

    International Nuclear Information System (INIS)

    Kohler, E.J.; Steward, K.P.; Iacono, J.V.

    1977-07-01

    The prime objective of the Peach Bottom End-of-Life Program was to validate specific HTGR design codes and predictions by comparison of actual and predicted physics, thermal, fission product, and materials behavior in Peach Bottom. Three consecutive phases of the program provide input to the HTGR design methods verifications: (1) Nondestructive fuel and circuit gamma scanning; (2) removal of steam generator and primary circuit components; and (3) Laboratory examinations of removed components. Component removal site work commenced with establishment of restricted access areas and installation of controlled atmosphere tents to retain relative humidity at <30%. A mock-up room was established to test and develop the tooling and to train operators under simulated working conditions. Primary circuit ducting samples were removed by trepanning, and steam generator access was achieved by a combination of arc gouging and grinding. Tubing samples were removed using internal cutters and external grinding. Throughout the component removal phase, strict health physics, safety, and quality assurance programs were implemented. A total of 148 samples of primary circuit ducting and steam generator tubing were removed with no significant health physics or safety incidents. Additionally, component removal served to provide access fordetermination of cesium plateout distribution by gamma scanning inside the ducts and for macroexamination of the steam generator from both the water and helium sides. Evaluations are continuing and indicate excellent performance of the steam generator and other materials, together with close correlation of observed and predicted fission product plateout distributions. It is concluded that such a program of end-of-life research, when appropriately coordinated with decommissioning activities, can significantly advance nuclear plant and fuel technology development

  9. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  10. Multipurpose nuclear process heat for energy supply in Brazil

    International Nuclear Information System (INIS)

    Hansen, U.; Inden, P.; Oesterwind, D.; Hukai, R.Y.; Pessine, R.T.; Pieroni, R.R.; Visoni, E.

    1978-11-01

    The industrialized nations require 75% of the energy as heat and it is likely that developing countries in the course of industrialization will show a comparable energy consumption structure. The High Temperature Reactor (HTR) allows the utilization of nuclear energy at high temperatures as process heat. In the Federal Republic of Germany (FRG) the development in the relevant technical areas is well advanced and warrants investigation as a matter for transfer to Brazil. In Brazil nuclear process heat finds possible applications in steel making, shale oil extraction, petroleum refining, and in the more distant future coal gasification with distribution networks. Based on growth forecasts for these industries a theoretical potential market of 38-53 GW (th) can be identified. At present nuclear process heat is marginally more expensive than conventional fossil technologies but the anticipated development is expected to add an economic incentive to the emerging necessity of providing a sound energy base in the developing countries. (author)

  11. Irreversibility and Action of the Heat Conduction Process

    Directory of Open Access Journals (Sweden)

    Yu-Chao Hua

    2018-03-01

    Full Text Available Irreversibility (that is, the “one-sidedness” of time of a physical process can be characterized by using Lyapunov functions in the modern theory of stability. In this theoretical framework, entropy and its production rate have been generally regarded as Lyapunov functions in order to measure the irreversibility of various physical processes. In fact, the Lyapunov function is not always unique. In the represent work, a rigorous proof is given that the entransy and its dissipation rate can also serve as Lyapunov functions associated with the irreversibility of the heat conduction process without the conversion between heat and work. In addition, the variation of the entransy dissipation rate can lead to Fourier’s heat conduction law, while the entropy production rate cannot. This shows that the entransy dissipation rate, rather than the entropy production rate, is the unique action for the heat conduction process, and can be used to establish the finite element method for the approximate solution of heat conduction problems and the optimization of heat transfer processes.

  12. Fundamentals of electroheat electrical technologies for process heating

    CERN Document Server

    Lupi, Sergio

    2017-01-01

    This book provides a comprehensive overview of the main electrical technologies for process heating, which tend to be treated separately in specialized books. Individual chapters focus on heat transfer, electromagnetic fields in electro-technologies, arc furnaces, resistance furnaces, direct resistance heating, induction heating, and high-frequency and microwave heating. The authors highlight those topics of greatest relevance to a wide-ranging teaching program, and at the same time offer a detailed review of the main applications of the various technologies. The content represents a synthesis of the extensive knowledge and experience that the authors have accumulated while researching and teaching at the University of Padua’s Engineering Faculty. This text on industrial electroheating technologies is a valuable resource not only for students of industrial, electrical, chemical, and material science engineering, but also for engineers, technicians and others involved in the application of electroheating and...

  13. Development of THYDE-HTGR: computer code for transient thermal-hydraulics of high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Hirano, Masashi; Hada, Kazuhiko

    1990-04-01

    The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)

  14. A new small HTGR power plant concept with inherently safe features--An engineering and economic challenge

    International Nuclear Information System (INIS)

    McDonald, C.F.; Sonn, D.L.

    1983-01-01

    This paper outlines a small nuclear plant concept which is not meant to replace the large nuclear power plants that will continue to be needed by the industrialized nations, but rather recognizes the needs of the smaller energy user, both for special applications in the US and for the developing nations. The small High-Temperature Gas-Cooled Reactor (HTGR), whose introduction will be very dependent on market forces, represents only one approach to meet these needs. The design of a small power plant that could be inherently safer and that might have costs less than those indicated by the traditional reverse-economy-of-scale effect is discussed. Topics considered include power plant economics, the small steam cycle HTGR thermodynamic cycle, the reactor nuclear heat source layout, the reactor heat removal system (main loop cooling, a vessel cooling system with reactor pressurized, vessel cooling system with reactor depressurized), safety considerations, investment risk protection, the technology base, and applications for the small HTGR plant concept

  15. Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, Friederike [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulation requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference

  16. 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.

  17. Pressurized Recuperator For Heat Recovery In Industrial High Temperature Processes

    Directory of Open Access Journals (Sweden)

    Gil S.

    2015-09-01

    Full Text Available Recuperators and regenerators are important devices for heat recovery systems in technological lines of industrial processes and should have high air preheating temperature, low flow resistance and a long service life. The use of heat recovery systems is particularly important in high-temperature industrial processes (especially in metallurgy where large amounts of thermal energy are lost to the environment. The article presents the process design for a high efficiency recuperator intended to work at high operating parameters: air pressure up to 1.2 MPa and temperature of heating up to 900°C. The results of thermal and gas-dynamic calculations were based on an algorithm developed for determination of the recuperation process parameters. The proposed technical solution of the recuperator and determined recuperation parameters ensure its operation under maximum temperature conditions.

  18. Prototype plant for nuclear process heat (PNP)

    International Nuclear Information System (INIS)

    Duerrfeld, R.; Kraut-Giesen, G.

    1982-01-01

    1. Goals: Verification of owner's interests during experimental and engineering phase of nuclear coal gasification. 2. Method: 2.1 Witnessing and evaluating of experimental results from running test facilities. 2.2 Influencing experimental program. 2.3 Participation in important meetings of PNP-project. 3. Results: From present point of view the realization of nuclear coal gasification with a nuclear high temperature reactor (HTR) in accordance with the present technical status as well as meeting the existing safety regulations seems to be feasable. R+D-work will be needed for affirmation of design. The gasification of hard coal basing on the allothermal principal has proved to be possible. The examination of the gasifier on a pilot scale is not yet done. The design work for the pilot plant should be started immediately, particularly keeping in mind the decision for erection of PNP in 1990. The calculation of production costs in comparison to autothermal gasification processes is promising better economics, if uncertainties of investment calculation are deemed to be neglectable. (orig.) [de

  19. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600 0 C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth

  20. Pre elementary design of primary reformer for hydrogen plant coupled with HTGR type NPP

    International Nuclear Information System (INIS)

    Dedy Priambodo; Erlan Dewita; Sudi Ariyanto

    2012-01-01

    Hydrogen has a high potent for new energy, because of it availability. Steam reforming is a fully developed commercial technology and is the most economical method for production of hydrogen. Steam reforming uses an external source of hot gas to heat tubes in which a catalytic reaction takes place that converts steam and lighter hydrocarbons such as natural gas (methane) or refinery feedstock into hydrogen and carbon monoxide (syngas) at high temperature on primary reformer (800-900°C). Utilization of helium from HTGR as heating medium for primary reformer has consequence to type and shape of its reactor. The main goal of this paper is to determine type/shape and pre elementary design of chemical reactor for the cogeneration system of Hydrogen Plant and HTGR The primary reformer for this system is Fixed Bed Multitube reactor with specification tube: NPS 3,5 Sch 40 ST 40S, 0.281 in thickness, number of tube 849 pieces and ASTM HH 30 for tube material. Tube arrangement is 'triangular pitch' on shell Split-Ring Floating Head from Steel Alloy SA 301 Grade B equipted with 8 baffles. (author)

  1. CHAP: a composite nuclear plant simulation program applied to the 3000 MW(t) HTGR

    International Nuclear Information System (INIS)

    Secker, P.A.; Bailey, P.G.; Gilbert, J.S.; Willcutt, G.J.E. Jr.; Vigil, J.C.

    1977-01-01

    The Composite HTGR Analysis Program (CHAP) is a general systems analysis program which has been developed at LASL. The program is being used for simulating large HTGR nuclear power plant operation and accident transients. The general features and analytical methods of the CHAP program are discussed. Features of the large HTGR model and results of model transients are also presented

  2. Energy analysis of a desalination process of sea water with nuclear energy; Analisis energetico de un proceso de desalinizacion de agua de mar con energia nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Martinez L, G.; Valle H, J., E-mail: julfi_jg@yahoo.com.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca No. 1009, Ex-hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2016-09-15

    In the present work, is theoretically proven that the residual heat, removed by the chillers in the stage prior to the compression of the recuperative Brayton cycle with which nuclear power plants operate with high temperature gas reactors (HTGR), can be used to produce stem and desalinate seawater. The desalination process selected for the analysis, based on its operating characteristics, is the Multi-Stage Distillation (Med). The Med process will use as energy source, for the flash evaporation process in the flash trap, the residual heat that the reactor coolant dissipates to the environment in order to increase the compression efficiency of the same; the energy dissipated depends on the operating conditions of the reactor. The Med distillation process requires saturated steam at low pressure which can be obtained by means of a heat exchanger, taking advantage of the residual heat, where the relative low temperatures with which the process operates make the nuclear plants with HTGR reactors ideal for desalination of sea water, because they do not require major modifications to their design of their operation. In this work the energy analysis of a six-stage Med module coupled to the chillers of an HTGR reactor of the Pebble Bed Modular Reactor type is presented. Mathematical modeling was obtained by differential equations of mass and energy balances in the system. The results of the analysis are presented in a table for each distillation stage, estimating the pure water obtained as a function of the heat supplied. (Author)

  3. A Simulation Study of Inter Heat Exchanger Process in SI Cycle Process for Hydrogen Production

    International Nuclear Information System (INIS)

    Shin, Jae Sun; Cho, Sung Jin; Choi, Suk Hoon; Qasim, Faraz; Lee, Euy Soo; Park, Sang Jin; Lee, Heung N.; Park, Jae Ho; Lee, Won Jae

    2014-01-01

    SI Cyclic process is one of the thermochemical hydrogen production processes using iodine and sulfur for producing hydrogen molecules from water. VHTR (Very High Temperature Reactor) can be used to supply heat to hydrogen production process, which is a high temperature nuclear reactor. IHX (Intermediate Heat Exchanger) is necessary to transfer heat to hydrogen production process safely without radioactivity. In this study, the strategy for the optimum design of IHX between SI hydrogen process and VHTR is proposed for various operating pressures of the reactor, and the different cooling fluids. Most economical efficiency of IHX is also proposed along with process conditions

  4. Match properties of heat transfer and coupled heat and mass transfer processes in air-conditioning system

    International Nuclear Information System (INIS)

    Zhang Tao; Liu Xiaohua; Zhang Lun; Jiang Yi

    2012-01-01

    Highlights: ► Investigates match properties of heat or mass transfer processes in HVAC system. ► Losses are caused by limited transfer ability, flow and parameter mismatching. ► Condition of flow matching is the same heat capacity of the fluids. ► Parameter matching is only reached along the saturation line in air–water system. ► Analytical solutions of heat and mass transfer resistance are derived. - Abstract: Sensible heat exchangers and coupled heat and mass transfer devices between humid air and water/desiccant are commonly used devices in air-conditioning systems. This paper focuses on the match properties of sensible heat transfer processes and coupled heat and mass transfer processes in an effort to understand the reasons for performance limitations in order to optimize system performance. Limited heat transfer capability and flow mismatching resulted in heat resistance of the sensible heat transfer process. Losses occurred during the heat and mass transfer processes due to limited transfer capability, flow mismatching, and parameter mismatching. Flow matching was achieved when the heat capacities of the fluids were identical, and parameter matching could only be reached along the saturation line in air–water systems or the iso-concentration line in air–desiccant systems. Analytical solutions of heat transfer resistance and mass transfer resistance were then derived. The heat and mass transfer process close to the saturation line is recommended, and heating sprayed water resulted in better humidification performance than heating inlet air in the air humidifier.

  5. Research of Snow-Melt Process on a Heated Platform

    Directory of Open Access Journals (Sweden)

    Vasilyev Gregory P.

    2016-01-01

    Full Text Available The article has shown the results of experimental researches of the snow-melt on a heated platform-near building heat-pump snow-melt platform. The near-building (yard heat pump platforms for snow melt with the area up to 10-15 m2 are a basis of the new ideology of organization of the street cleaning of Moscow from snow in the winter period which supposes the creation in the megalopolis of the «distributed snow-melt system» (DSMS using non-traditional energy sources. The results of natural experimental researches are presented for the estimation of efficiency of application in the climatic conditions of Moscow of heat pumps in the snow-melt systems. The researches were conducted on a model sample of the near-building heat-pump platform which uses the low-potential thermal energy of atmospheric air. The conducted researches have confirmed experimentally in the natural conditions the possibility and efficiency of using of atmospheric air as a source of low-potential thermal energy for evaporation of the snow-melt heat pump systems in the climatic conditions of Moscow. The results of laboratory researches of snow-melt process on a heated horizontal platform are presented. The researches have revealed a considerable dependence of efficiency of the snow-melt process on its piling mode (form-building and the organization of the process of its piling mode (form-building and the organization of the process of its (snow mass heat exchange with the surface of the heated platform. In the process of researches the effect of formation of an «ice dome» under the melting snow mass called by the fact that in case of the thickness of snow loaded on the platform more than 10 cm the water formed from the melting snow while the contact with the heating surface don’t spread on it, but soaks into the snow, wets it due to capillary effect and freezes. The formation of «ice dome» leads to a sharp increase of snow-melt period and decreases the operating

  6. Cogeneration using a nuclear reactor to generate process heat

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon

    2009-01-01

    Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

  7. 1170-MW(t) HTGR-PS/C plant application study report: shale oil recovery application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    The US has large shale oil energy resources, and many companies have undertaken considerable effort to develop economical means to extract this oil within environmental constraints. The recoverable shale oil reserves in the US amount to 160 x 10 9 m 3 (1000 x 10 9 bbl) and are second in quantity only to coal. This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to a shale oil recovery process. Since the highest potential shale oil reserves lie in th Piceance Basin of Western Colorado, the study centers on exploiting shale oil in this region

  8. Induction heating in in-line strip production process

    International Nuclear Information System (INIS)

    Costa, P.; Santinelli, M.

    1995-05-01

    ISP (In-line Strip Production), a continuous process for steel strip production, has recently been set in an italian innovative plant, where ecological impact and power requirements are lighter than usual. This report describes the studies performed by ENEA (Italian Agency for New Technologies, Energy and the Environment), while a prototype reheating facility was arranged by Acciaieria ISP in Cremona (Italy). The authors, after a study of the prototype electromagnetic field, calculate the heating rate, with the thermal network method. Then they detect, with a 1-D-FEM, the heat diffusion through the strip cross section. Afterward, since the heat distribution depends on the eddy current density one, which is given by the magnetic field distribution, the authors, with a 3-D-FEM, carry out a coupled, electromagnetic and thermal, analysis in time domain. The strip temperature map is established by the balance between skin depth heating and surface cooling: a thermal analysis, performed with a moving 2-D-FEM, take into account the effects of the different heating and cooling situations, originated by the strip moving at a speed of 6m/min through four consecutive reheating facilities. The temperatures of a strip sample heated by the prototype have been monitored, acquired by a computer and related with the simulation results. The little difference between experiment and simulation assessed the qualitative and quantitative validity of this analysis, that has come out to be a tool, useful to evaluate the effects of possible improvements to the ISP process

  9. Progress Report for Diffusion Welding of the NGNP Process Application Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    R.E. Mizia; D.E. Clark; M.V. Glazoff; T.E. Lister; T.L. Trowbridge

    2011-04-01

    The NGNP Project is currently investigating the use of metallic, diffusion welded, compact heat exchangers to transfer heat from the primary (reactor side) heat transport system to the secondary heat transport system. The intermediate heat exchanger will transfer this heat to downstream applications such as hydrogen production, process heat, and electricity generation. The channeled plates that make up the heat transfer surfaces of the intermediate heat exchanger will have to be assembled into an array by diffusion welding.

  10. Design and evaluation of heat utilization systems for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    2001-08-01

    The primary focus of this CRP was to perform detailed investigation of the high temperature industrial processes that are attainable through incorporation of an HTGR, and for their possible demonstration in the HTTR. The HTGR has the capability to achieve a core outlet temperature approaching 1,000 deg. C in a safe and effective manner. These attributes, coupled with the offer by JAERI to utilize the HTTR, resulted in the initiation of this CRP by the IAEA. High Temperature Engineering Test Reactor (HTTR) utilizes a 30 MW(th) HTGR comprised of 30 fuel columns of hexagonal pin-in-pin graphite block type fuel elements. The fuel consists of UO 2 TRISO coated particles with an enrichment of ∼ 6% wt. Relative to the demonstration of high temperature heat applications, the HTTR will be capable of producing 10 MW(th) of heat at 950 deg. C. However, the thermal power for these applications has the potential to be increased up to 30 MW(th) in the future, which may be required for demonstration of gas turbine system components. The HTTR reached initial criticality in November 1998. Initial operational plans includes a series of rise to power tests followed by tests to demonstrate the safety and operational characteristics of the HTTR. In addition to completion of the HTTR demonstration tests, it was recommended that the R and D be performed within the HTTR project. JAERI is encouraged to publicize the results of the HTTR tests and 'lessons learned' from their experiences including potential capabilities of the HTGR for heat applications. The next priority application was determined to be the generation of electricity through the use of the gas turbine. Application of the Brayton Cycle utilizing high temperature helium from a modular HTGR was chosen for development because of its projected benefits as an economic and efficient means for the production of electricity. Evaluation of the remaining high temperature heat utilization applications chosen for investigation resulted

  11. Subharmonic excitation in an HTGR core

    International Nuclear Information System (INIS)

    Bezler, P.; Curreri, J.R.

    1977-01-01

    The occurrence of subharmonic resonance in a series of blocks with clearance between blocks and with springs on the outer most ends is the subject of this paper. This represents an HTGR core response to an earthquake input. An analytical model of the cross section of this type of core is a series of blocks arranged horizontally between outer walls. Each block represents many graphite hexagonal core elements acting in unison as a single mass. The blocks are of unequal size to model the true mass distribution through the core. Core element elasticity and damping characteristics are modeled with linear spring and viscous damping units affixed to each block. The walls and base represent the core barell or core element containment structure. For forced response calculations, these boundaries are given prescribed motions. The clearance between each block could be the same or different with the total clearance duplicating that of the entire core. Spring packs installed between the first and last block and the boundaries model the boundary elasticity. The system non-linearity is due to the severe discontinuity in the interblock elastic forces when adjacent blocks collide. A computer program using a numerical integration scheme was developed to solve for the response of the system to arbitrary inputs

  12. High-temperature Gas Reactor (HTGR)

    Science.gov (United States)

    Abedi, Sajad

    2011-05-01

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

  13. Numerical simulation of plasma processes driven by transverse ion heating

    Science.gov (United States)

    Singh, Nagendra; Chan, C. B.

    1993-01-01

    The plasma processes driven by transverse ion heating in a diverging flux tube are investigated with numerical simulation. The heating is found to drive a host of plasma processes, in addition to the well-known phenomenon of ion conics. The downward electric field near the reverse shock generates a doublestreaming situation consisting of two upflowing ion populations with different average flow velocities. The electric field in the reverse shock region is modulated by the ion-ion instability driven by the multistreaming ions. The oscillating fields in this region have the possibility of heating electrons. These results from the simulations are compared with results from a previous study based on a hydrodynamical model. Effects of spatial resolutions provided by simulations on the evolution of the plasma are discussed.

  14. A pilot test plan of the thermochemical water-splitting iodine-sulfur process

    International Nuclear Information System (INIS)

    Kubo, Shinji; Kasahara, Seiji; Okuda, Hiroyuki; Terada, Atsuhiko; Tanaka, Nobuyuki; Inaba, Yoshitomo; Ohashi, Hirofumi; Inagaki, Yoshiyuki; Onuki, Kaoru; Hino, Ryutaro

    2004-01-01

    Research and development (R and D) of hydrogen production systems using high-temperature gas-cooled reactors (HTGR) are being conducted by the Japan Atomic Research Institute (JAERI). To develop the systems, superior hydrogen production methods are essential. The thermochemical hydrogen production cycle, the IS (iodine-sulfur) process, is a prospective candidate, in which heat supplied by HTGR can be consumed for the thermal driving load. With this attractive feature, JAERI will conduct pilot-scale tests, aiming to establish technical bases for practical plant designs using HTGR. The hydrogen will be produced at a maximum rate of 30 m 3 /h, continuously using high-temperature helium gas supplied by a helium gas loop, with an electric heater of about 400 kW. The plant will employ an advanced hydroiodic acid-processing device for efficient hydrogen production, and the usefulness of the device was confirmed from mass and heat balance analysis. Through design works and the hydrogen production tests, valuable data for construction and operation will be acquired to evaluate detailed process performance for practical systems. After completing the pilot-scale tests, JAERI will move onto the next R and D step, which will be demonstrations of the IS process to which heat is supplied from a high-temperature engineering test reactor (HTTR)

  15. Characterization and processing of heat treated aluminium matrix composite

    Science.gov (United States)

    Doifode, Yogesh; Kulkarni, S. G.

    2018-05-01

    The present study is carried out to determine density and porosity of Aluminium bagasse ash reinforced composite produced by powder metallurgy method. Bagasse ash is used as reinforcement material having high silica and alumina contents and varied from 5 weight % to 40 weight%. The manufactured composite is heat treated, the main objective of heat treatment is to prepare the material structurally and physically fit for engineering application. The results showed that the density decreases with percentage increase in reinforcement of bagasse ash from 2.6618 gm/cm3 to 1.9830 gm/cm3 with the minimum value at 40 weight% bagasse ash without heat treatment whereas after heat treatment density of composite increases due filling up of voids and porous holes. Heat treatment processing is the key to this improvement, with the T6 heat treated composite to convene the reduced porosity of composite. Consequently aluminium metal matrix composite combines the strength of the reinforcement to achieve a combination of desirable properties not available in any single material. It may observe that porosity in case of powder metallurgy samples showed more porosity portions compare to the casting samples. In order to achieve optimality in structure and properties of Bagasse ash-reinforcement heat treatment techniques have evolved. Generally, the ceramic reinforcements increase the density of the base alloy during fabrication of composites. However, the addition of lightweight reinforcements reduces the density of the hybrid composites. The results also showed that, the density varies from to with minimum value at 40 wt. % BA. The results of the statistical analysis showed that there are significant differences among the means of each property of the composites at various levels of BA replacement .It was concluded that bagasse ash can be used as reinforcement and the produced composites have low density and heat treatment reduces porosity which could be used in automobile industry for

  16. Trace Metal Levels in Raw and Heat Processed Nigerian Staple ...

    African Journals Online (AJOL)

    The levels of some trace metals (Fe, Zn, Cu, Ni, Cd) were quantitatively determined in raw and heat processed staple food cultivars (yam, cassava, cocoyam and maize) from oil producing areas of part of the Niger Delta and compared with a non-oil producing area of Ebonyi State as control. The survey was conducted to ...

  17. Plug and Play Process Control of a District Heating System

    DEFF Research Database (Denmark)

    Trangbaek, Klaus; Knudsen, Torben; Skovmose Kallesøe, Carsten

    2009-01-01

    The main idea of plug and play process control is to initialise and reconfigure control systems automatically. In this paper these ideas are applied to a scaled laboratory model of a district heating pressure control system.  First of all this serves as a concrete example of plug and play control...

  18. Weldability of general purpose heat source new-process iridium

    International Nuclear Information System (INIS)

    Kanne, W.R.

    1987-01-01

    Weldability tests on General Purpose Heat Source (GPHS) iridium capsules showed that a new iridium fabrication process reduced susceptibility to underbead cracking. Seventeen capsules were welded (a total of 255 welds) in four categories and the number of cracks in each weld was measured

  19. r-PROCESS LANTHANIDE PRODUCTION AND HEATING RATES IN KILONOVAE

    Energy Technology Data Exchange (ETDEWEB)

    Lippuner, Jonas; Roberts, Luke F., E-mail: jlippuner@tapir.caltech.edu [TAPIR, Walter Burke Institute for Theoretical Physics, California Institute of Technology, MC 350-17, 1200 E California Boulevard, Pasadena CA 91125 (United States)

    2015-12-20

    r-process nucleosynthesis in material ejected during neutron star mergers may lead to radioactively powered transients called kilonovae. The timescale and peak luminosity of these transients depend on the composition of the ejecta, which determines the local heating rate from nuclear decays and the opacity. Kasen et al. and Tanaka and Hotokezaka pointed out that lanthanides can drastically increase the opacity in these outflows. We use the new general-purpose nuclear reaction network SkyNet to carry out a parameter study of r-process nucleosynthesis for a range of initial electron fractions Y{sub e}, initial specific entropies s, and expansion timescales τ. We find that the ejecta is lanthanide-free for Y{sub e} ≳ 0.22−0.30, depending on s and τ. The heating rate is insensitive to s and τ, but certain, larger values of Y{sub e} lead to reduced heating rates, due to individual nuclides dominating the heating. We calculate approximate light curves with a simplified gray radiative transport scheme. The light curves peak at about a day (week) in the lanthanide-free (-rich) cases. The heating rate does not change much as the ejecta becomes lanthanide-free with increasing Y{sub e}, but the light-curve peak becomes about an order of magnitude brighter because it peaks much earlier when the heating rate is larger. We also provide parametric fits for the heating rates between 0.1 and 100 days, and we provide a simple fit in Y{sub e}, s, and τ to estimate whether or not the ejecta is lanthanide-rich.

  20. r-PROCESS LANTHANIDE PRODUCTION AND HEATING RATES IN KILONOVAE

    International Nuclear Information System (INIS)

    Lippuner, Jonas; Roberts, Luke F.

    2015-01-01

    r-process nucleosynthesis in material ejected during neutron star mergers may lead to radioactively powered transients called kilonovae. The timescale and peak luminosity of these transients depend on the composition of the ejecta, which determines the local heating rate from nuclear decays and the opacity. Kasen et al. and Tanaka and Hotokezaka pointed out that lanthanides can drastically increase the opacity in these outflows. We use the new general-purpose nuclear reaction network SkyNet to carry out a parameter study of r-process nucleosynthesis for a range of initial electron fractions Y e , initial specific entropies s, and expansion timescales τ. We find that the ejecta is lanthanide-free for Y e ≳ 0.22−0.30, depending on s and τ. The heating rate is insensitive to s and τ, but certain, larger values of Y e lead to reduced heating rates, due to individual nuclides dominating the heating. We calculate approximate light curves with a simplified gray radiative transport scheme. The light curves peak at about a day (week) in the lanthanide-free (-rich) cases. The heating rate does not change much as the ejecta becomes lanthanide-free with increasing Y e , but the light-curve peak becomes about an order of magnitude brighter because it peaks much earlier when the heating rate is larger. We also provide parametric fits for the heating rates between 0.1 and 100 days, and we provide a simple fit in Y e , s, and τ to estimate whether or not the ejecta is lanthanide-rich

  1. HTGR-GT and electrical load integrated control

    International Nuclear Information System (INIS)

    Chan, T.; Openshaw, F.; Pfremmer, D.

    1980-05-01

    A discussion of the control and operation of the HTGR-GT power plant is presented in terms of its closely coupled electrical load and core cooling functions. The system and its controls are briefly described and comparisons are made with more conventional plants. The results of analyses of selected transients are presented to illustrate the operation and control of the HTGR-GT. The events presented were specifically chosen to show the controllability of the plant and to highlight some of the unique characteristics inherent in this multiloop closed-cycle plant

  2. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  3. HTGR containment design options: an application of probabilistic risk assessment

    International Nuclear Information System (INIS)

    1977-08-01

    Through the use of probabilistic risk assessment (PRA), it is possible to quantitatively evaluate the radiological risk associated with a given reactor design and to place such risk into perspective with alternative designs. The merits are discussed for several containment alternatives for the HTGR from the viewpoints of economics and licensability, as well as public risk. The quantification of cost savings and public risk indicates that presently acceptable public risk can be maintained and cost savings of $40 million can result from use of a vented confinement for the HTGR

  4. HTGR structural-materials efforts in the US

    International Nuclear Information System (INIS)

    Rittenhouse, P.L.; Roberts, D.I.

    1982-07-01

    The status of ongoing structural materials programs being conducted in the US to support development and deployment of the high-temperature gas-cooled reactor (HTGR) is described. While the total US program includes work in support of all variants of this reactor system, the emphasis of this paper is on the work aimed at support of the steam cycle/cogeneration (SC/C) version of the HTGR. Work described includes activities to develop design and performance prediction data on metals, ceramics, and graphite

  5. Process heat exchanger for SO3 decomposer fabricated with Ni-based alloys surface modified by SiC film deposition and N ion beam bombardment

    International Nuclear Information System (INIS)

    Park, Jae-Won; Kim, Hyung-Jin; Choi, Yong-Woon; Kim, Yong-Wan

    2007-01-01

    In the iodine-sulfur (IS) cycle for the hydrogen production using the high temperature gas-cooled reactor (HTGR), one of the important components is the SO 3 decomposer which generates SO 2 and SO 3 gases under high temperature conditions. Since this environment is extremely corrosive, the materials used for the decomposer should meet excellent mechanical properties at the elevated temperature as well as high corrosion resistance in SO 2 /SO 3 atmospheres. In general, ceramics are protective against the corrosion, but metals exhibit limited corrosion resistance. In this work, the ceramic coating on the metallic substrate was studied. We selected SiC as coating materials and Ni-based alloys as the substrate materials. Since the adhesion between the coated layer and the substrate is most crucial in this application, we attempted to develop Ion Beam Mixing (IBM) technique to produce a highly adherent coated layer. For the fabrication of process heat exchange for SO 3 decomposer, the diffusion bonding at ∼900 .deg. C is employed because this temperature does not affect the mechanical properties of materials

  6. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    International Nuclear Information System (INIS)

    Tang, I.N.; Munkelwitz, H.R.

    1982-07-01

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated

  7. Study on reprocessing of uranium-thorium fuel with solvent extraction for HTGR

    International Nuclear Information System (INIS)

    Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun

    1992-08-01

    A single cycle process by solvent extraction with acid feed solution is suggested. The purpose is to reprocess uranium-thorium fuel elements which are of high burn-up and rich of 232 U from HTGR (high temperature gas cooled reactor). The extraction cascade tests have been completed. The recovery of uranium and thorium is greater than 99.6%. By this method, the requirement, under remote control to re-fabricate fuel elements, of decontamination factors for Cs, Sr, Zr-Nb and Ru has been reached

  8. 1170-MW(t) HTGR-PS/C plant application study report: tar sands oil recovery application

    International Nuclear Information System (INIS)

    Rao, R.; McMain, A.T. Jr.

    1981-05-01

    This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to tar sands oil recovery and upgrading. The raw product recovered from the sands is a heavy, sour bitumen; upgrading, which involves coking and hydrodesulfurization, produces a synthetic crude (refinable by current technology) and petroleum coke. Steam and electric power are required for the recovery and upgrading process. Proposed and commercial plants would purchase electric power from local utilities and obtain from boilers fired with coal and with by-product fuels produced by the upgrading. This study shows that an HTGR-PS/C represents a more economical source of steam and electric power

  9. Present status of research on and development of HTGR techniques in the People's Republic of China

    International Nuclear Information System (INIS)

    Zhu Yongjun

    1989-01-01

    China is a developing country rich in coal, petroleum and hydropower resources. In the past ten years, energy production in China has had a large increase, but along with the development of economy, energy demands increase even more rapidly. Many problems exist in China's energy system. Considering the large energy demand in the near future and long-term energy strategy, China has already decided to develop nuclear power gradually. The first several nuclear power stations are being and will be built in the South-east sea shore region. Two 900 MW PWRs (from France) and one 300 MW PWR (home made) are now under construction at Daya Bay (Kwangton Province) and Qin Shan (Zhejiang Province). The succeeding PWR power plants are being planned. PWR nuclear power station has been selected for the beginning of China's nuclear power plan. For large scale utilization of nuclear power in the next century, the development of advanced reactor type with good safety and economy performances and high uranium utilization rate (uranium resources in China is not rich enough) is strategically important. HTGR, due to its inherent safety characteristics, high heat efficiency, flexible fuel system and wide application fields, is a prospective advanced reactor type. Research and development on HTGR have already been included in China's national technical development program and are going on smoothly

  10. 9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Canning with heat processing and hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... IDENTIFICATION AS TO CLASS, QUALITY, QUANTITY, AND CONDITION Inspection Procedure § 355.25 Canning with heat...

  11. Shape Effect on the Temperature Field during Microwave Heating Process

    Directory of Open Access Journals (Sweden)

    Zhijun Zhang

    2018-01-01

    Full Text Available Aiming at improving the food quality during microwave process, this article mainly focused on the numerical simulation of shape effect, which was evaluated by microwave power absorption capability and temperature distribution uniformity in a single sample heated in a domestic microwave oven. This article only took the electromagnetic field and heat conduction in solid into consideration. The Maxwell equations were used to calculate the distribution of microwave electromagnetic field distribution in the microwave cavity and samples; then the electromagnetic energy was coupled as the heat source in the heat conduction process in samples. Quantitatively, the power absorption capability and temperature distribution uniformity were, respectively, described by power absorption efficiency (PAE and the statistical variation of coefficient (COV. In addition, we defined the comprehensive evaluation coefficient (CEC to describe the usability of a specific sample. In accordance with volume or the wave numbers and penetration numbers in the radial and axial directions of samples, they can be classified into different groups. And according to the PAE, COV, and CEC value and the specific need of microwave process, an optimal sample shape and orientation could be decided.

  12. Fuel production from coal by the Mobil Oil process using nuclear high-temperature process heat

    International Nuclear Information System (INIS)

    Hoffmann, G.

    1982-01-01

    Two processes for the production of liquid hydrocarbons are presented: Direct conversion of coal into fuel (coal hydrogenation) and indirect conversion of coal into fuel (syngas production, methanol synthesis, Mobil Oil process). Both processes have several variants in which nuclear process heat may be used; in most cases, the nuclear heat is introduced in the gas production stage. The following gas production processes are compared: LURGI coal gasification process; steam reformer methanation, with and without coal hydrogasification and steam gasification of coal. (orig./EF) [de

  13. Heat and work distributions for mixed Gauss–Cauchy process

    International Nuclear Information System (INIS)

    Kuśmierz, Łukasz; Gudowska-Nowak, Ewa; Rubi, J Miguel

    2014-01-01

    We analyze energetics of a non-Gaussian process described by a stochastic differential equation of the Langevin type. The process represents a paradigmatic model of a nonequilibrium system subject to thermal fluctuations and additional external noise, with both sources of perturbations considered as additive and statistically independent forcings. We define thermodynamic quantities for trajectories of the process and analyze contributions to mechanical work and heat. As a working example we consider a particle subjected to a drag force and two statistically independent Lévy white noises with stability indices α = 2 and α = 1. The fluctuations of dissipated energy (heat) and distribution of work performed by the force acting on the system are addressed by examining contributions of Cauchy fluctuations (α = 1) to either bath or external force acting on the system. (paper)

  14. Numerical Analysis of Heat Transfer During Quenching Process

    Science.gov (United States)

    Madireddi, Sowjanya; Krishnan, Krishnan Nambudiripad; Reddy, Ammana Satyanarayana

    2018-04-01

    A numerical model is developed to simulate the immersion quenching process of metals. The time of quench plays an important role if the process involves a defined step quenching schedule to obtain the desired characteristics. Lumped heat capacity analysis used for this purpose requires the value of heat transfer coefficient, whose evaluation requires large experimental data. Experimentation on a sample work piece may not represent the actual component which may vary in dimension. A Fluid-Structure interaction technique with a coupled interface between the solid (metal) and liquid (quenchant) is used for the simulations. Initial times of quenching shows boiling heat transfer phenomenon with high values of heat transfer coefficients (5000-2.5 × 105 W/m2K). Shape of the work piece with equal dimension shows less influence on the cooling rate Non-uniformity in hardness at the sharp corners can be reduced by rounding off the edges. For a square piece of 20 mm thickness, with 3 mm fillet radius, this difference is reduced by 73 %. The model can be used for any metal-quenchant combination to obtain time-temperature data without the necessity of experimentation.

  15. Heat and mass transfer enhancement in absorbing processes

    International Nuclear Information System (INIS)

    Hijikata, Kunio; Lee, S.K.

    1993-01-01

    The key to improving the performance of absorption-type heat machines lies in the enhancement of the mass transfer of the vapor into the absorbant solution, since the mass diffusivity in the solution is very small compared to the thermal diffusivity. The absorption process is influenced by many factors including physical properties of the fluids, the flow pattern and others, especially the velocity profile near the interface is the most important. From these stand points, the heat and mass transfer in the absorption was investigated by following three steps. First, an augmentation of the absorption to a liquid film flowing in groove was theoretically investigated, in which the interface between the vapor and liquid film is cooled by the grooved surfaces. Secondly, systematical experiments were carried out on several factors that affect the absorption process, which were the cooling wall temperature, the inlet solution subcooling, and the fin configuration. Finally, a numerical study of the heat and mass transfer enhancement due to flow agitation by the periodically grooved channel was conducted. That flow realized by fabricating ridges on the fin surface. A secondary flow due to these ridges is expected to enhance the heat and mass transfer. These results were compared with experimental ones. (orig.)

  16. Biodiesel production process from microalgae oil by waste heat recovery and process integration.

    Science.gov (United States)

    Song, Chunfeng; Chen, Guanyi; Ji, Na; Liu, Qingling; Kansha, Yasuki; Tsutsumi, Atsushi

    2015-10-01

    In this work, the optimization of microalgae oil (MO) based biodiesel production process is carried out by waste heat recovery and process integration. The exergy analysis of each heat exchanger presented an efficient heat coupling between hot and cold streams, thus minimizing the total exergy destruction. Simulation results showed that the unit production cost of optimized process is 0.592$/L biodiesel, and approximately 0.172$/L biodiesel can be avoided by heat integration. Although the capital cost of the optimized biodiesel production process increased 32.5% and 23.5% compared to the reference cases, the operational cost can be reduced by approximately 22.5% and 41.6%. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  18. Safety and licensing analyses for the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.; Harrington, R.M.; Cleveland, J.C.; Clapp, N.E. Jr.

    1982-01-01

    The Oak Ridge National Laboratory (ORNL) safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks during a postulated accident

  19. Proceedings of the 1st JAERI symposium on HTGR technologies

    International Nuclear Information System (INIS)

    1990-07-01

    This report was edited as the Proceedings of the 1st JAERI Symposium on HTGR Technologies, - Design, Licensing Requirements and Supporting Technologies -, collecting the 21 papers presented in the Symposium. The 19 of the presented papers are indexed individually. (J.P.N.)

  20. Gasification of coal using nuclear process heat. Chapter D

    International Nuclear Information System (INIS)

    Schilling, H.-D.; Bonn, B.; Krauss, U.

    1979-01-01

    In the light of the high price of coal and the enormous advances made recently in nuclear engineering, the possibility of using heat from high-temperature nuclear reactors for gasification processes was discussed as early as the 1960s. The advantages of this technology are summarized. A joint programme of development work is described, in which the Nuclear Research Centre at Juelich is aiming to develop a high-temperature reactor which will supply process heat at as high a temperature as possible, while other organizations are working on the hydrogasification of lignites and hard coals, and steam gasification. Experiments are at present being carried out on a semi-technical scale, and no operational data for large-scale plants are available as yet. (author)

  1. Overall simulation of a HTGR plant with the gas adapted MANTA code

    International Nuclear Information System (INIS)

    Emmanuel Jouet; Dominique Petit; Robert Martin

    2005-01-01

    Full text of publication follows: AREVA's subsidiary Framatome ANP is developing a Very High Temperature Reactor nuclear heat source that can be used for electricity generation as well as cogeneration including hydrogen production. The selected product has an indirect cycle architecture which is easily adapted to all possible uses of the nuclear heat source. The coupling to the applications is implemented through an Intermediate Heat exchanger. The system code chosen to calculate the steady-state and transient behaviour of the plant is based on the MANTA code. The flexible and modular MANTA code that is originally a system code for all non LOCA PWR plant transients, has been the subject of new developments to simulate all the forced convection transients of a nuclear plant with a gas cooled High Temperature Reactor including specific core thermal hydraulics and neutronics modelizations, gas and water steam turbomachinery and control structure. The gas adapted MANTA code version is now able to model a total HTGR plant with a direct Brayton cycle as well as indirect cycles. To validate these new developments, a modelization with the MANTA code of a real plant with direct Brayton cycle has been performed and steady-states and transients compared with recorded thermal hydraulic measures. Finally a comparison with the RELAP5 code has been done regarding transient calculations of the AREVA indirect cycle HTR project plant. Moreover to improve the user-friendliness in order to use MANTA as a systems conception, optimization design tool as well as a plant simulation tool, a Man- Machine-Interface is available. Acronyms: MANTA Modular Advanced Neutronic and Thermal hydraulic Analysis; HTGR High Temperature Gas-Cooled Reactor. (authors)

  2. Conceptual design of small-sized HTGR system (4). Plant design and technical feasibility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing L.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio

    2013-09-01

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2020s. HTR50S was designed for steam supply and electricity generation by the steam turbine with the reactor outlet temperature of 750degC as a reference plant configuration. On the other hand, the intermediate heat exchanger (IHX) will be installed in the primary loop to demonstrate the electricity generation by the helium gas turbine and hydrogen production by thermochemical water splitting by utilizing the secondary helium loop with the reactor outlet temperature of 900degC as a future plant configuration. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750degC to 900degC and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems (i.e., reactor internal components, reactor pressure vessel, vessel cooling system, shutdown cooling system, steam generator (SG), gas circulator, SG isolation and drainage system, reactor containment vessel, steam turbine and heat supply system) satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S. (author)

  3. Influence of heat processing methods on the nutrient composition ...

    African Journals Online (AJOL)

    No significant difference (P > 0.05) was obtained in the saponification number of the three samples analyzed and values ranged from 161.3 ±2.92 in RS to 163.0 ± 2.60 in FS. Heat processing (boiling and frying) generally decreased significantly (p<0.05) the crude protein, crude fat, caloric value, Fe, Zn, vitamins A and C as ...

  4. Potential industrial market for process heat from nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, R.W.

    1976-07-01

    A specific segment of industrial process heat use has been examined in detail to identify individual plant locations throughout the United states where nuclear generated steam may be a viable alternative. Five major industries have been studied: paper, chemicals, petroleum, rubber, and primary metals. For these industries, representing 75 percent of the total industrial steam consumption, the individual plant locations within the U.S. using steam in large quantities have been located and characterized as to fuel requirements

  5. Preparation of silicon carbide nanowires via a rapid heating process

    International Nuclear Information System (INIS)

    Li Xintong; Chen Xiaohong; Song Huaihe

    2011-01-01

    Silicon carbide (SiC) nanowires were fabricated in a large quantity by a rapid heating carbothermal reduction of a novel resorcinol-formaldehyde (RF)/SiO 2 hybrid aerogel in this study. SiC nanowires were grown at 1500 deg. C for 2 h in an argon atmosphere without any catalyst via vapor-solid (V-S) process. The β-SiC nanowires were characterized by field-emission scanning electron microscope (FE-SEM), X-ray diffraction (XRD), transmission electron microscope (TEM), high-resolution transmission electron microscope (HRTEM) equipped with energy dispersive X-ray (EDX) facility, Fourier transformed infrared spectroscopy (FTIR), and thermogravimetric analysis (TGA). The analysis results show that the aspect ratio of the SiC nanowires via the rapid heating process is much larger than that of the sample produced via gradual heating process. The SiC nanowires are single crystalline β-SiC phase with diameters of about 20-80 nm and lengths of about several tens of micrometers, growing along the [1 1 1] direction with a fringe spacing of 0.25 nm. The role of the interpenetrating network of RF/SiO 2 hybrid aerogel in the carbothermal reduction was discussed and the possible growth mechanism of the nanowires is analyzed.

  6. Public acceptance of HTGR technology - HTR2008-58218

    International Nuclear Information System (INIS)

    Hannink, R.; Kuhr, R.; Morris, T.

    2008-01-01

    Nuclear energy projects continue to evoke strong emotional responses from the general public throughout the world. High Temperature Gas-Cooled Reactor (HTGR) technology offers improved safety and performance characteristics that should enhance public acceptance but is burdened with demonstrating a different set of safety principles. This paper summarizes key issues impacting public acceptance and discusses the importance of openly engaging the public in the early stages of new HTGR projects. The public gets information about new technologies through schools and universities, news and entertainment media, the internet, and other forms of information exchange. Development of open public forums, access to information in understandable formats, participation of universities in preparing and distributing educational materials, and other measures will be needed to support widespread public confidence in the improved safety and performance characteristics of HTGR technology. This confidence will become more important as real projects evolve and participants from outside the nuclear industry begin to evaluate the real and perceived risks, including potential impacts on public relations, branding, and shareholder value when projects are announced. Public acceptance and support will rely on an informed understanding of the issues and benefits associated with HTGR technology. Major issues of public concern include nuclear safety, avoidance of greenhouse gas emissions, depletion of natural gas resources, energy security, nuclear waste management, local employment and economic development, energy prices, and nuclear proliferation. Universities, the media, private industry, government entities, and other organizations will all have roles that impact public acceptance, which will likely play a critical role in the future markets, siting, and permitting of HTGR projects. (authors)

  7. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  8. Process Design of Aluminum Tailor Heat Treated Blanks

    Directory of Open Access Journals (Sweden)

    Alexander Kahrimanidis

    2015-12-01

    Full Text Available In many industrials field, especially in the automotive sector, there is a trend toward lightweight constructions in order to reduce the weight and thereby the CO2 and NOx emissions of the products. An auspicious approach within this context is the substitution of conventional deep drawing steel by precipitation hardenable aluminum alloys. However, based on the low formability, the application for complex stamping parts is challenging. Therefore, at the Institute of Manufacturing Technology, an innovative technology to enhance the forming limit of these lightweight materials was invented. The key idea of the so-called Tailor Heat Treated Blanks (THTB is optimization of the mechanical properties by local heat treatment before the forming operation. An accurate description of material properties is crucial to predict the forming behavior of tailor heat treated blanks by simulation. Therefore, within in this research project, a holistic approach for the design of the THTB process in dependency of the main influencing parameters is presented and discussed in detail. The capability of the approach for the process development of complex forming operations is demonstrated by a comparison of local blank thickness of a tailgate with the corresponding results from simulation.

  9. Control of the tritium path in process heat HTR's

    International Nuclear Information System (INIS)

    Kirch, N.; Scheidler, G.

    1985-01-01

    Nuclear Process Heat plant converting fossil fuels into liquid or gaseous secondary energy carriers generate tritium by several nuclear reactions. Control of the tritium path through the walls of the heat exchanger is highly important to meet regulatory requirements on the acceptable contamination in the product gas or liquid. Therefore, significant effort in the project 'Prototypanlage Nukleare Prozesswaerme' was put not only into generating a data base, but also into means of reducing tritium generation and permeation. Clean graphites with lithium impurities in the ppb level provide a low tritium source term. Realistic modeling of graphite retention and special helium purification systems are essentials. The main barrier to tritium permeation are heat exchanger walls requiring detailed characterization of in-situ surface layers. Studies to optimize the water/steam mass flow in the conversion process offer possibilities for further tritium retention. Progress can be demonstrated as follows: In 1980, between 2 and 8 Bq tritium per gram of product were predicted based on available data and even higher concentrations during startup. However, present day validated code predictions are below required 0.5 Bq/g equilibrium concentration level. During transients - particularly startup - this limit cannot be guaranteed as yet, but further retention potential is being offered by tritium gettering or filtering. An expected increase of the German regulatory requirement to 5 Bq/g will easily be met by present plant design under all operational conditions. (author)

  10. Particle Acceleration and Heating Processes at the Dayside Magnetopause

    Science.gov (United States)

    Berchem, J.; Lapenta, G.; Richard, R. L.; El-Alaoui, M.; Walker, R. J.; Schriver, D.

    2017-12-01

    It is well established that electrons and ions are accelerated and heated during magnetic reconnection at the dayside magnetopause. However, a detailed description of the actual physical mechanisms driving these processes and where they are operating is still incomplete. Many basic mechanisms are known to accelerate particles, including resonant wave-particle interactions as well as stochastic, Fermi, and betatron acceleration. In addition, acceleration and heating processes can occur over different scales. We have carried out kinetic simulations to investigate the mechanisms by which electrons and ions are accelerated and heated at the dayside magnetopause. The simulation model uses the results of global magnetohydrodynamic (MHD) simulations to set the initial state and the evolving boundary conditions of fully kinetic implicit particle-in-cell (iPic3D) simulations for different solar wind and interplanetary magnetic field conditions. This approach allows us to include large domains both in space and energy. In particular, some of these regional simulations include both the magnetopause and bow shock in the kinetic domain, encompassing range of particle energies from a few eV in the solar wind to keV in the magnetospheric boundary layer. We analyze the results of the iPic3D simulations by discussing wave spectra and particle velocity distribution functions observed in the different regions of the simulation domain, as well as using large-scale kinetic (LSK) computations to follow particles' time histories. We discuss the relevance of our results by comparing them with local observations by the MMS spacecraft.

  11. Influence of microwave heating on the stability of processed samn

    Directory of Open Access Journals (Sweden)

    Farag, Radwan S.

    1991-04-01

    Full Text Available Butter was converted to samn by microwave and conventional heating. The quality of the processed samn by the two methods was followed by determining the acid, peroxide and TBA values over a period of six weeks at 60°C. The fatty acid composition of samn samples was determined by gas-liquid chromatographic technique. The data show that butter conversion to samn by microwave heating was accomplished in about one half of the time that conventional heating requires. Microwave heating obviously increased the development of samn rancidity compared with the conventional heating. The parameters used for measuring lipid rancidity indicated that the main cause of samn rancidity under the present conditions is an oxidation mechanism.

    Mantequilla fue transformada en samn por calentamiento en microonda y convencional. La calidad del elaborado de samn por los dos métodos fue seguida mediante determinación de los índices de acidez, peróxido y TBA durante un período de seis semanas a 60°C. La composición en ácidos grasos de muestras de samn fue determinada por técnica cromatográfica gas-líquido. Los datos mostraron que la conversión de mantequilla a samn por calentamiento en microonda fue realizada en aproximadamente una vez y media el tiempo que exige el calentamiento convencional. El calentamiento en microonda, evidentemente, aumentó el desarrollo de la rancidez del samn comparado con el calentamiento convencional. Los parámetros usados para la medida de la rancidez lipídica indicaron que la causa principal de la rancidez del samn bajo las condiciones presentes es un mecanismo de oxidación.

  12. SONATINA-2V: a computer program for seismic analysis of the two-dimensional vertical slice HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1982-07-01

    A computer program SONATINA-2V has been developed for predicting the behavior of a two-dimensional vertical slice HTGR core under seismic excitation. SONATINA-2V is a general two-dimensional computer program capable of analyzing the vertical slice HTGR core with the permanent side reflector blocks and its restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Coulomb friction is taken into account between blocks and between dowel pin and hole. A spring dashpot model is used for the collision process between adjacent blocks. The core support structure is represented by a single block. The computer program SONATINA-2V is capable of analyzing the core behavior for an excitation input applied simultaneously to both vertical and horizontal directions. Analytical results obtained from SONATINA-2V are compared with experimental results and are found to be in good agreement. The computer program can thus be used to predict with a good accuracy the behavior of the HTGR core under seismic excitation. In the present report are given, the theoretical formulation of the analytical model, a user's manual to describe the input and output format, and sample problems. (author)

  13. Experimental determination of thermal conductivity and gap conductance of fuel rod for HTGR

    International Nuclear Information System (INIS)

    Kikuchi, Teruo; Iwamoto, Kazumi; Ikawa, Katsuichi; Ishimoto, Kiyoshi

    1985-01-01

    The thermal conductivity of fuel compacts and the gap conductance between the fuel compact and the graphite sleeve in fuel rods for a high-temperature gas-cooled reactor (HTGR) were measured by the center heating method. These measurements were made as functions of volume percent particle loading and temperature for thermal conductivity and as functions of gap distance and gas composition for gap conductance. The thermal conductivity of fuel compacts decreases with increasing temperature and with increasing particle loading. The gap conductance increases with increasing temperature and decrease with increasing gap distance. A good gap conductance was observed with helium fill gas. It was seen that the gap conductance was dependent on the thermal conductivity of fill gas and conductance by radiation and could be neglected the conductance through solid-solid contact points of fuel compact and graphite sleeve. (author)

  14. HTGR molten salt sensible energy transmission and storage system design and costs

    International Nuclear Information System (INIS)

    1981-09-01

    This report, which was prepared for Gas-Cooled Reactor Associates by United Engineers and Constructors under Contract No. GCRA/UE and C 81-203, presents the design and cost for a molten salt Sensible Energy Transmission and Storage (SETS) System. Although the reference system for this study is sized to be compatible with an 1170 MW(t) HTGR Nuclear Heat Source, the results and conclusions should be generally applicable to most large scale molten salt energy transmission system applications. A preliminary conceptual design is presented and alternative configurations are discussed. The sensitivity of system costs to variations in important system parameters are also presented. Costs for a reference case conceptual design are reported in constant 1980 dollars and inflated 1995 dollars

  15. Digital simulation of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant

    International Nuclear Information System (INIS)

    Ray, A.; Bowman, H.F.

    1978-01-01

    A nonlinear dynamic model of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant was derived in state-space form from fundamental principles. The plant model is 40th order, time-invariant, deterministic and continuous-time. Numerical results were obtained by digital simulation. Steady-state performance of the nonlinear model was verified with plant heat balance data at 100, 75 and 50 percent load levels. Local stability, controllability and observability were examined in this range using standard linear algorithms. Transfer function matrices for the linearized models were also obtained. Transient response characteristics of 6 system variables for independent step distrubances in 2 different input variables are presented as typical results

  16. Improved process for heating finely divided carbonaceous materials

    Energy Technology Data Exchange (ETDEWEB)

    1956-08-01

    A process for heating finely divided carbonaceous particles by burning a proportion of the carbon consists of passing the carbonaceous material at a temperature above 800/sup 0/F into an upwardly disposed, slender, combustion zone, suspending the particles in an upwardly-moving gas containing free oxygen so that the suspension has a density from 0.1 to 5.0 lb/cu. ft., passing the suspension upwardly through the combustion zone at a velocity of from 5 to 100 ft./sec., and injecting at least one stream of a second gas containing free oxygen at a point in the combustion zone such that at least 50% of the oxygen in the first gas has been consumed by the time the suspension reaches this point. The total quantity of oxygen is chosen so that the finely divided carbonaceous material is heated to a temperature of not less than 1,050/sup 0/F.

  17. Students’ Conception on Heat and Temperature toward Science Process Skill

    Science.gov (United States)

    Ratnasari, D.; Sukarmin, S.; Suparmi, S.; Aminah, N. S.

    2017-09-01

    This research is aimed to analyze the effect of students’ conception toward science process skill. This is a descriptive research with subjects of the research were 10th-grade students in Surakarta from high, medium and low categorized school. The sample selection uses purposive sampling technique based on physics score in national examination four latest years. Data in this research collecting from essay test, two-tier multiple choice test, and interview. Two-tier multiple choice test consists of 30 question that contains an indicator of science process skill. Based on the result of the research and analysis, it shows that students’ conception of heat and temperature affect science process skill of students. The students’ conception that still contains the wrong concept can emerge misconception. For the future research, it is suggested to improve students’ conceptual understanding and students’ science process skill with appropriate learning method and assessment instrument because heat and temperature is one of physics material that closely related with students’ daily life.

  18. Measurement of Weight of Kernels in a Simulated Cylindrical Fuel Compact for HTGR

    International Nuclear Information System (INIS)

    Kim, Woong Ki; Lee, Young Woo; Kim, Young Min; Kim, Yeon Ku; Eom, Sung Ho; Jeong, Kyung Chai; Cho, Moon Sung; Cho, Hyo Jin; Kim, Joo Hee

    2011-01-01

    The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. The weight of fuel kernels in an element is generally measured by the chemical analysis or a gamma-ray spectrometer. Although it is accurate to measure the weight of kernels by the chemical analysis, the samples used in the analysis cannot be put again in the fabrication process. Furthermore, radioactive wastes are generated during the inspection procedure. The gamma-ray spectrometer requires an elaborate reference sample to reduce measurement errors induced from the different geometric shape of test sample from that of reference sample. X-ray computed tomography (CT) is an alternative to measure the weight of kernels in a compact nondestructively. In this study, X-ray CT is applied to measure the weight of kernels in a cylindrical compact containing simulated TRISO-coated particles with ZrO 2 kernels. The volume of kernels as well as the number of kernels in the simulated compact is measured from the 3-D density information. The weight of kernels was calculated from the volume of kernels or the number of kernels. Also, the weight of kernels was measured by extracting the kernels from a compact to review the result of the X-ray CT application

  19. The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

    1979-01-01

    The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

  20. Simulating the heat transfer process of horizontal anode baking furnace

    Energy Technology Data Exchange (ETDEWEB)

    L.Q. Zhang; C.G. Zheng; M.H. Xu [Huazhong University of Science and Technology, Wuhan (China). State Key Laboratory of Coal Combustion

    2005-07-01

    A transient two-dimensional mathematical model of a horizontal baking furnace is presented. The model combines complex thermal phenomena in a baking process such as air infiltration, evolution and combustion of volatile matters, combustion of packing coke, and heat losses. The predicted results are in good agreement with measured data. Furthermore, the process is simulated under different operating conditions such as firing cycle time, airflow and air infiltration. The simulated results indicate that the fuel consumption decreases as the firing cycle time decreases. It is also found that reducing the airflow and air infiltration will help to save fuel. The model is proved to be a useful tool for the process optimisation of the baking furnace in the aluminum industry.

  1. Research of processes of heat exchange in horizontal pipeline

    Science.gov (United States)

    Nikolaev, A. K.; Dokoukin, V. P.; Lykov, Y. V.; Fetisov, V. G.

    2018-03-01

    The energy crisis, which becomes more evident in Russia, stems in many respects from unjustified high consumption of energy resources. Development and exploitation of principal oil and gas deposits located in remote areas with severe climatic conditions require considerable investments increasing essentially the cost of power generation. Account should be taken also of the fact that oil and gas resources are nonrenewable. An alternative fuel for heat and power generation is coal, the reserves of which in Russia are quite substantial. For this reason the coal extraction by 2020 will amount to 450-550 million tons. The use of coal, as a solid fuel for heat power plants and heating plants, is complicated by its transportation from extraction to processing and consumption sites. Remoteness of the principal coal mining areas (Kuzbass, Kansk-Achinsk field, Vorkuta) from the main centers of its consumption in the European part of the country, Siberia and Far East makes the problem of coal transportation urgent. Of all possible transportation methods (railway, conveyor, pipeline), the most efficient is hydrotransport which provides continuous transportation at comparatively low capital and working costs, as confirmed by construction and operation of extended coal pipelines in many countries.

  2. Utilization of geothermal heat in tropical fruit-drying process

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B.H.; Lopez, L.P.; King, R.; Fujii, J.; Tanaka, M.

    1982-10-01

    The power plant utilizes only the steam portion of the HGP-A well production. There are approximately 50,000 pounds per hour of 360/sup 0/F water produced (approximately 10 million Btu per hour) and the water is currently not used and is considered a waste. This tremendous resource could very well be used in applications such as food processing, food dehydration and other industrial processing that requires low-grade heat. One of the applications is examined, namely the drying of tropical fruits particularly the papaya. The papaya was chosen for the obvious reason that it is the biggest crop of all fruits produced on the Big Island. A conceptual design of a pilot plant facility capable of processing 1000 pounds of raw papaya per day is included. This facility is designed to provide a geothermally heated dryer to dehydrate papayas or other tropical fruits available on an experimental basis to obtain data such as drying time, optimum drying temperature, etc.

  3. Prediction of deformations of steel plate by artificial neural network in forming process with induction heating

    International Nuclear Information System (INIS)

    Nguyen, Truong Thinh; Yang, Young Soo; Bae, Kang Yul; Choi, Sung Nam

    2009-01-01

    To control a heat source easily in the forming process of steel plate with heating, the electro-magnetic induction process has been used as a substitute of the flame heating process. However, only few studies have analyzed the deformation of a workpiece in the induction heating process by using a mathematical model. This is mainly due to the difficulty of modeling the heat flux from the inductor traveling on the conductive plate during the induction process. In this study, the heat flux distribution over a steel plate during the induction process is first analyzed by a numerical method with the assumption that the process is in a quasi-stationary state around the inductor and also that the heat flux itself greatly depends on the temperature of the workpiece. With the heat flux, heat flow and thermo-mechanical analyses on the plate to obtain deformations during the heating process are then performed with a commercial FEM program for 34 combinations of heating parameters. An artificial neural network is proposed to build a simplified relationship between deformations and heating parameters that can be easily utilized to predict deformations of steel plate with a wide range of heating parameters in the heating process. After its architecture is optimized, the artificial neural network is trained with the deformations obtained from the FEM analyses as outputs and the related heating parameters as inputs. The predicted outputs from the neural network are compared with those of the experiments and the numerical results. They are in good agreement

  4. Industrial and agricultural process heat information user study

    Energy Technology Data Exchange (ETDEWEB)

    Belew, W.W.; Wood, B.L.; Marle, T.L.; Reinhardt, C.L.

    1981-03-01

    The results of a series of telephone interviews with groups of users of information on solar industrial and agricultural process heat (IAPH) are described. These results, part of a larger study on many different solar technologies, identify types of information each group needed and the best ways to get information to each group. In the current study only high-priority groups were examined. Results from 10 IAPH groups of respondents are analyzed in this report: IPH Researchers; APH Researchers; Representatives of Manufacturers of Concentrating and Nonconcentrating Collectors; Plant, Industrial, and Agricultural Engineers; Educators; Representatives of State Agricultural Offices; and County Extension Agents.

  5. Synthesis of hydrocarbons using coal and nuclear process heat

    International Nuclear Information System (INIS)

    Eickhoff, H.G.; Kugeler, K.

    1975-01-01

    An analysis of the global petroleum resources and demand shows that the amount of mineral oil products is sufficient to meet the requirements of the next decades. The geographical resources, however, could lead to problems of distribution and foreign exchange. The production of hydrocarbons with coal as basis using high temperature nuclear process heat has advantages compared to the conventional techniques. Next to the conservation of reserve fossil primary energy carriers there are advantages as regards prices, which at high coal costs are especially pronounced. (orig.) [de

  6. Innovative food processing technology using ohmic heating and aseptic packaging for meat.

    Science.gov (United States)

    Ito, Ruri; Fukuoka, Mika; Hamada-Sato, Naoko

    2014-02-01

    Since the Tohoku earthquake, there is much interest in processed foods, which can be stored for long periods at room temperature. Retort heating is one of the main technologies employed for producing it. We developed the innovative food processing technology, which supersede retort, using ohmic heating and aseptic packaging. Electrical heating involves the application of alternating voltage to food. Compared with retort heating, which uses a heat transfer medium, ohmic heating allows for high heating efficiency and rapid heating. In this paper we ohmically heated chicken breast samples and conducted various tests on the heated samples. The measurement results of water content, IMP, and glutamic acid suggest that the quality of the ohmically heated samples was similar or superior to that of the retort-heated samples. Furthermore, based on the monitoring of these samples, it was observed that sample quality did not deteriorate during storage. © 2013. Published by Elsevier Ltd on behalf of The American Meat Science Association. All rights reserved.

  7. Manufacturing of tailored tubes with a process integrated heat treatment

    Science.gov (United States)

    Hordych, Illia; Boiarkin, Viacheslav; Rodman, Dmytro; Nürnberger, Florian

    2017-10-01

    The usage of work-pieces with tailored properties allows for reducing costs and materials. One example are tailored tubes that can be used as end parts e.g. in the automotive industry or in domestic applications as well as semi-finished products for subsequent controlled deformation processes. An innovative technology to manufacture tubes is roll forming with a subsequent inductive heating and adapted quenching to obtain tailored properties in the longitudinal direction. This processing offers a great potential for the production of tubes with a wide range of properties, although this novel approach still requires a suited process design. Based on experimental data, a process simulation is being developed. The simulation shall be suitable for a virtual design of the tubes and allows for gaining a deeper understanding of the required processing. The model proposed shall predict microstructural and mechanical tube properties by considering process parameters, different geometries, batch-related influences etc. A validation is carried out using experimental data of tubes manufactured from various steel grades.

  8. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  9. Evaluation of the significance of inverse oxidation for HTGR graphites

    International Nuclear Information System (INIS)

    Lee, B.S.; Heiser, J. III; Sastre, C.

    1983-01-01

    The inverse oxidation refers to a higher mass loss inside the graphite than the outside. In 1980, Wichner et al reported this phenomenon (referred to as inside/out corrosion) observed in some H451 graphites, and offered an explanation that a catalyst (almost certainly Fe) is activated by the progressively increasing reducing conditions found in the graphite interior. Recently, Morgan and Thomas (1982) investigated this phenomenon is PGX graphites, and agreed on the existing mechanism to explain this pheomenon. They also called for attention to the possibility that this phenomenon may occur under HTGR (High Temperature Gas-Cooled Reactor) operating conditions. The purpose of this paper is to confirm the above mentioned explanation for this phenomenon and to evaluate the significance of this effect for HTGR graphites under realistic reactor conditions

  10. Energy conversion processes for the use of geothermal heat

    Energy Technology Data Exchange (ETDEWEB)

    Minder, R. [Minder Energy Consulting, Oberlunkhofen (Switzerland); Koedel, J.; Schaedle, K.-H.; Ramsel, K. [Gruneko AG, Basel (Switzerland); Girardin, L.; Marechal, F. [Swiss Federal Institute of Technology (EPFL), Laboratory for industrial energy systems (LENI), Lausanne (Switzerland)

    2007-03-15

    This comprehensive final report for the Swiss Federal Office of Energy (SFOE) presents the results of a study made on energy conversion processes that can be used when geothermal heat is to be used. The study deals with both theoretical and practical aspects of the conversion of geothermal heat to electricity. The report is divided into several parts and covers general study, practical experience, planning and operation of geothermal power plants as well as methodology for the optimal integration of energy conversion systems in geothermal power plants. In the first part, the specific properties and characteristics of geothermal resources are discussed. Also, a general survey of conversion processes is presented with special emphasis on thermo-electric conversion. The second part deals with practical aspects related to planning, construction and operation of geothermal power plant. Technical basics, such as relevant site-specific conditions, drilling techniques, thermal water or brine quality and materials requirements. Further, planning procedures are discussed. Also, operation and maintenance aspects are examined and some basic information on costs is presented. The third part of the report presents the methodology and results for the optimal valorisation of the thermodynamic potential of deep geothermal systems.

  11. Prototype plant for nuclear process heat (PNP), reference phase

    International Nuclear Information System (INIS)

    Fladerer, R.; Schrader, L.

    1982-07-01

    The coal gasification processes using nuclear process heat being developed within the framwork of the PNP project, have the advantages of saving feed coal, improving efficiency, reducing emissions, and stabilizing energy costs. One major gasification process is the hydrogasification of coal for producing SNG or gas mixture of carbon monoxide and hydrogen; this process can also be applied in a conventional route. The first steps to develop this process were planning, construction and operation of a semi-technical pilot plant for hydrogasification of coal in a fluidized bed having an input of 100 kg C/h. Before the completion of the development phase (reference phase) describing here, several components were tested on part of which no operational experience had so far been gained; these were the newly developed devices, e.g. the inclined tube for feeding coal into the fluidized bed, and the raw gas/hydrogenation gas heat exchanger for utilizing the waste heat of the raw gas leaving the gasifier. Concept optimizing of the thoroughly tested equipment parts led to an improved operational behaviour. Between 1976 and 1980, the semi-technical pilot plant was operated for about 19,400 hours under test conditions, more than 7,400 hours of which it has worked under gasification conditions. During this time approx. 1,100 metric tons of dry brown coal and more than 13 metric tons of hard coal were gasified. The longest coherent operational phase under gasification conditions was 748 hours in which 85.4 metric tons of dry brown coal were gasified. Carbon gasification rates up to 82% and methane contents in the dry raw gas (free of N 2 ) up to 48 vol.% were obtained. A detailed evaluation of the test results provided information of the results obtained previously. For the completion of the test - primarily of long-term tests - the operation of the semi-technical pilot plant for hydrogasification of coal is to be continued up to September 1982. (orig.) [de

  12. Steam gasification of coal, project prototype plant nuclear process heat

    International Nuclear Information System (INIS)

    Heek, K.H. van

    1982-05-01

    This report describes the tasks, which Bergbau-Forschung has carried out in the field of steam gasification of coal in cooperation with partners and contractors during the reference phase of the project. On the basis of the status achieved to date it can be stated, that the mode of operation of the gas-generator developed including the direct feeding of caking high volatile coal is technically feasible. Moreover through-put can be improved by 65% at minimum by using catalysts. On the whole industrial application of steam gasification - WKV - using nuclear process heat stays attractive compared with other gasification processes. Not only coal is conserved but also the costs of the gas manufactured are favourable. As confirmed by recent economic calculations these are 20 to 25% lower. (orig.) [de

  13. A Fresnel collector process heat experiment at Capitol Concrete Products

    Science.gov (United States)

    Hauger, J. S.

    1981-01-01

    An experiment is planned, conducted and evaluated to determine the feasibility of using a Power Kinetics' Fresnel concentrator to provide process heat in an industrial environment. The plant provides process steam at 50 to 60 psig to two autoclaves for curing masonry blocks. When steam is not required, the plant preheats hot water for later use. A second system is installed at the Jet Propulsion Laboratory parabolic dish test site for hardware validation and experiment control. Experiment design allows for the extrapolation of results to varying demands for steam and hot water, and includes a consideration of some socio-technical factors such as the impact on production scheduling of diurnal variations in energy availability.

  14. Scaling laws for HTGR core block seismic response

    International Nuclear Information System (INIS)

    Dove, R.C.

    1977-01-01

    This paper discusses the development of scaling laws, physical modeling, and seismic testing of a model designed to represent a High Temperature Gas-Cooled Reactor (HTGR) core consisting of graphite blocks. The establishment of the proper scale relationships for length, time, force, and other parameters is emphasized. Tests to select model materials and the appropriate scales are described. Preliminary results obtained from both model and prototype systems tested under simulated seismic vibration are presented

  15. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  16. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    Shatoff, H.; Charman, C.M.

    1983-01-01

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

  17. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  18. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  19. HTGR safety research at the Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Stroh, K.R.; Anderson, C.A.; Kirk, W.L.

    1982-01-01

    This paper summarizes activities undertaken at the Los Alamos National Laboratory as part of the High-Temperature Gas-Cooled Reactor (HTGR) Safety Research Program sponsored by the US Nuclear Regulatory Commission. Technical accomplishments and analysis capabilities in six broad-based task areas are described. These tasks are: fission-product technology, primary-coolant impurities, structural investigations, safety instrumentation and control systems, accident delineation, and phenomena modeling and systems analysis

  20. Study of air ingress accident of an HTGR

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1995-01-01

    Inherent properties of high temperature gas cooled reactors (HTGR) facilitate the design of HTGRs with high degree of passive safety performances. In this context, it is very important to establish a design criteria for a passive safe function for the air ingress accident. However, it is absolutely necessary to investigate the air ingress behavior during the accident before exploring the design criteria. The present paper briefly describes major activities and results of the air ingress research in our laboratory. (author)

  1. Nuclear district heating. 1. Process heat reactors and transmission and distribution networks

    International Nuclear Information System (INIS)

    Caizergues, R.

    1979-01-01

    Three kinds of production station are considered: joint electricity and heat-producing stations, heat-producing stations with CAS reactors and heat-producing stations with Thermos reactors. The thermal energy supply possibilities of these stations, the cost price of this energy and the cost price per therm produced by the district heating source and conveyed to the user are studied [fr

  2. GTOROTO: a simulation system for HTGR core seismic behavior

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Nakamura, Yasuhiro; Onuma, Yoshio

    1980-07-01

    One of the most important design of HTGR core is its aseismic structure. Therefore, it is necessary to predict the forces and motion of the core blocks. To meet the requirement, many efforts to develop analytical methods and computer programs are made. A graphic simulation system GTOROTO with a CRT graphic display and lightpen was developed to analyze the HTGR core behavior in seismic excitation. Feature of the GTOROTO are as follows: (1) Behavior of the block-type HTGR core during earthquake can be shown on the CRT-display. (2) Parameters of the computing scheme can be changed with the lightpen. (3) Routines of the computing scheme can be changed with the lightpen and an alteration switch. (4) Simulation pictures are shown automatically. Hardcopies are available by plotter in stopping the progress of simulation pictures. Graphic representation can be re-start with the predetermined program. (5) Graphic representation informations can be stored in assembly language on a disk for rapid representation. (6) A computer-generated cinema can be made by COM (Computer Output Microfilming) or filming directly the CRT pictures. These features in the GTOROTO are provided in on-line conversational mode. (author)

  3. Management feature of transuranic for HTGR and LWR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Long-lived actinides from spent fuels can cause potential long-term environ- mental hazards. The generation and incineration of transuranic in different closed fuel cycles were studied. U and Pu were recycled from spent fuel in the 250 MW high-temperature gas-cooled reactor-pebble-bed-module (HTR-PM) U-Pu fuelled core, and then PuO 2 and MOX fuel elements were designed based on this recycled U and Pu. These fuel elements were used to build up a new PuO 2 or MOX fuelled core with the same geometry of the original reactor. Characteristics of transuranic incineration with HTGR open and closed fuel cycles were studied with VSOP code, and the corresponding results from the light water reactor were compared and analyzed. The transuranic generation with HTGR open fuel cycle is almost half of the corresponding result of the light water reactor. Thus, HTGR closed fuel cycles can effectively burn transuranic. (authors)

  4. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  5. Status of the HTGR development program in Japan

    International Nuclear Information System (INIS)

    Saito, S.

    1991-01-01

    According to the revision of the Long-Term Program for Development and Utilization of Nuclear Energy issued by the Japanese Atomic Energy Commission, High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, will be constructed by the Japan Atomic Energy Research Institute (JAERI) in order to establish and upgrade the technology basis for an HTGR, serving at the same time as a potential tool for new and innovative basic research. The budget for the construction of the HTTR was approved by the Government and JAERI is now proceeding with the construction design of the HTTR, focussing the first criticality in the end of FY 1995. In order to establish and upgrade HTGR technology basis systematically and efficiently, and also to carry out innovative basic research on high temperature technologies, Japan will perform necessary R and D mainly at JAERI, which is a leading organization of the R and D. In addition, in order to promote the R and D on HTGRs more efficiently, Japan will promote the existing international cooperation with the research organizations in foreign countries. (author). 5 figs, 3 tabs

  6. Large deviations in stochastic heat-conduction processes provide a gradient-flow structure for heat conduction

    International Nuclear Information System (INIS)

    Peletier, Mark A.; Redig, Frank; Vafayi, Kiamars

    2014-01-01

    We consider three one-dimensional continuous-time Markov processes on a lattice, each of which models the conduction of heat: the family of Brownian Energy Processes with parameter m (BEP(m)), a Generalized Brownian Energy Process, and the Kipnis-Marchioro-Presutti (KMP) process. The hydrodynamic limit of each of these three processes is a parabolic equation, the linear heat equation in the case of the BEP(m) and the KMP, and a nonlinear heat equation for the Generalized Brownian Energy Process with parameter a (GBEP(a)). We prove the hydrodynamic limit rigorously for the BEP(m), and give a formal derivation for the GBEP(a). We then formally derive the pathwise large-deviation rate functional for the empirical measure of the three processes. These rate functionals imply gradient-flow structures for the limiting linear and nonlinear heat equations. We contrast these gradient-flow structures with those for processes describing the diffusion of mass, most importantly the class of Wasserstein gradient-flow systems. The linear and nonlinear heat-equation gradient-flow structures are each driven by entropy terms of the form −log ρ; they involve dissipation or mobility terms of order ρ 2 for the linear heat equation, and a nonlinear function of ρ for the nonlinear heat equation

  7. Structural integrity assessment of intermediate heat exchanger in the HTTR. Based on results of rise-to-power test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-12-01

    A helium/helium intermediate heat exchanger (IHX) in the high temperature engineering test reactor (HTTR) is an essential component for demonstration of future nuclear process heat utilization of high temperature gas-cooled reactor (HTGR). The IHX with a heat capacity of 10 MW has 96 helically-coiled heat transfer tubes. Structural design for the IHX had been conducted through elastic-creep analysis of superalloy Hastelloy XR components such as heat transfer tubes and center pipe. In the HTTR rise-to-power test, it was clarified that temperature of the coolant in the IHX at the reactor scrams changes more rapidly than expected in the design. Effects of the IHX coolant temperature change, at anticipated reactor scram from the full power of 30 MW at high temperature test operation, on structural integrity of the heat transfer tubes and the lower reducer of the center pipe were investigated analytically based on the coolant temperature data obtained from the rise-to-power test. As results of the assessment, it was confirmed that cumulative principal creep strain, cumulative creep and fatigue damage factor of the IHX components during 10{sup 5} h of the HTTR lifetime should be below the allowable limits, which are established in the high-temperature structural design code for the HTGR Class 1 components. (author)

  8. The maximum power condition of the brayton cycle with heat exchange processes

    International Nuclear Information System (INIS)

    Jung, Pyung Suk; Cha, Jin Girl; Ro, Sung Tack

    1985-01-01

    The ideal brayton cycle has been analyzed with the heat exchange processes between the working fluid and the heat source and the sink while their heat capacity rates are constant. The power of the cycle can be expressed in terms of a temperature of the cycle and the heat capacity rate of the working fluid. There exists an optimum power condition where the heat capacity rate of the working fluid has a value between those of the heat source and the heat sink, and the cycle efficiency is determined by the inlet temperatures of the heat source and the sink. (Author)

  9. High Magnetic Field Processing - A Heat-Free Heat Treating Method

    Energy Technology Data Exchange (ETDEWEB)

    Ludtka, Gerard Michael [ORNL; Ludtka, Gail Mackiewicz- [ORNL; Wilgen, John B [ORNL; Kenik, Edward A [ORNL; Parish, Chad M [ORNL; Rios, Orlando [ORNL; Rogers, Hiram [ORNL; Manuel, Michele [University of Florida, Gainesville; Kisner, Roger A [ORNL; Watkins, Thomas R [ORNL; Murphy, Bart L [ORNL

    2012-08-01

    The High and Thermal Magnetic Processing/Electro-magnetic Acoustic Transducer (HTMP/EMAT) technology has been shown to be an enabling disruptive materials processing technology, that can achieve significant improvements in microstructure and consequently material performance beyond that achievable through conventional processing, and will lead to the next generation of advanced performance structural and functional materials. HTMP exposure increased the reaction kinetics enabling refinement of microstructural features such as finer martensite lath size, and finer, more copious, homogeneous dispersions of strengthening carbides leading to combined strength and toughness improvements in bainitic steels. When induction heating is applied in a high magnetic field environment, the induction heating coil is configured so that high intensity acoustic/ultrasonic treatment occurs naturally. The configuration results in a highly effective electromagnetic acoustical transducer (EMAT). HTMP combined with applying high-field EMAT, produce a non-contact ultrasonic treatment that can be used to process metal alloys in either the liquid state resulting in significant microstructural changes over conventional processing. Proof-of-principle experiments on cast irons resulted in homogeneous microstructures in small castings along with improved casting surface appearance. The experiment showed that by exposing liquid metal to the non-contact acoustic/ultrasonic processing technology developed using HMFP/EMAT wrought-like microstructures were developed in cast components. This Energy Intensive Processes (EIP) project sponsored by the DOE EERE Advanced Manufacturing Office (AMO) demonstrated the following: (1) The reduction of retained austenite in high carbon/high alloy steels with an ambient temperature HTMP process, replacing either a cryogenic or double tempering thermal process normally employed to accomplish retained austenite transformation. HTMP can be described as a 'heat

  10. A numerical study of EGS heat extraction process based on a thermal non-equilibrium model for heat transfer in subsurface porous heat reservoir

    Science.gov (United States)

    Chen, Jiliang; Jiang, Fangming

    2016-02-01

    With a previously developed numerical model, we perform a detailed study of the heat extraction process in enhanced or engineered geothermal system (EGS). This model takes the EGS subsurface heat reservoir as an equivalent porous medium while it considers local thermal non-equilibrium between the rock matrix and the fluid flowing in the fractured rock mass. The application of local thermal non-equilibrium model highlights the temperature-difference heat exchange process occurring in EGS reservoirs, enabling a better understanding of the involved heat extraction process. The simulation results unravel the mechanism of preferential flow or short-circuit flow forming in homogeneously fractured reservoirs of different permeability values. EGS performance, e.g. production temperature and lifetime, is found to be tightly related to the flow pattern in the reservoir. Thermal compensation from rocks surrounding the reservoir contributes little heat to the heat transmission fluid if the operation time of an EGS is shorter than 15 years. We find as well the local thermal equilibrium model generally overestimates EGS performance and for an EGS with better heat exchange conditions in the heat reservoir, the heat extraction process acts more like the local thermal equilibrium process.

  11. Approach to the HTGR core outlet temperature measurements in the United States

    International Nuclear Information System (INIS)

    Franklin, R.; Rodriguez, C.

    1982-06-01

    The High Temperature Gas-Cooled Reactor (HTGR) constructed at Fort St. Vrain Colorado (330 MWe) used Geminol thermocouples to measure the primary coolant temperature at the core outlet. The primary coolant (helium) is heated by the graphite core to temperatures in the range of 700 deg. to 750 deg. C. The combination of the high temperature, high flow rate and radiation at the core outlet area makes it difficult to obtain accurate temperature measurements. The Geminol thermocouples installed in the Fort St. Vrain reactor have provided accurate data for several years of power operation without any failures. The indicated temperature of the core outlet thermocouples agrees with a ''traversing'' thermocouple measurement to within +-2 deg. C. The Geminol thermocouple wire was provided by the Driver-Harris Company and is similar to the chromel versus alumel thermocouple. Geminol wire is no longer distributed and on future designs, chromel versus alumel wire will be used. The next large HTGR design, which is being performed with funding support from the United States Department of Energy, will incorporate replaceable thermocouples. The thermocouples used in the Fort St. Vrain reactor were permanently installed and large in diameter (6.35 mm) to insure good reliability. The replaceable thermocouples to be used in the next large reactor will be smaller in diameter (3.18 mm). These replaceable thermocouples will be inserted into the core outlet area through long curved guide tubes that are permanently installed. These guide tubes are as long as 18 meters and must be curved to reach the core outlet regions. Tests were conducted to prove that the thermocouples could be inserted and removed through the long curved guide tubes. (author)

  12. System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

    International Nuclear Information System (INIS)

    McKellar, Michael G.; Harvego, Edwin A.; Gandrik, Anastasia A.

    2010-01-01

    A design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322 C and 750 C, respectively. The power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with 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 cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.

  13. Soil Heat Flow. Physical Processes in Terrestrial and Aquatic Ecosystems, Transport Processes.

    Science.gov (United States)

    Simpson, James R.

    These materials were designed to be used by life science students for instruction in the application of physical theory to ecosystem operation. Most modules contain computer programs which are built around a particular application of a physical process. Soil heat flow and the resulting soil temperature distributions have ecological consequences…

  14. Ways to increase efficiency of the HTGR coupled with the gas-turbine power conversion unit - HTR2008-58274

    International Nuclear Information System (INIS)

    Golovko, V. F.; Kodochigov, N. G.; Vasyaev, A. V.; Shenoy, A.; Baxi, C. B.

    2008-01-01

    The paper deals with the issue of increasing efficiency of nuclear power plants with the modular high-temperature helium reactor (HTGR) and direct gas turbine cycle. It should be noted that only this combination can highlight the advantages of the HTGR, namely the ability to heat helium to about 1000 deg. C, in comparison with other reactor plants for electricity generation. The HTGR has never been used in the direct gas turbine cycle. At present, several designs of such commercial plants are at the stage of experimental validation of main technical features. In Russia, 'OKB Mechanical Engineering' together with 'General Atomics' (USA) are developing the GT-MHR project with the reactor power of 600 MW, reactor outlet helium temperature of 850 deg. C, and efficiency of about 45.2%; the South African Republic is developing the PBMR project with the reactor power of 400 MW, reactor outlet helium temperature of 900 deg. C, and efficiency of about 42%; and Japan is developing the GTHTR-300 project with the reactor power of 600 MW, reactor outlet helium temperature of 850 deg. C, and efficiency of about 45.6%. As it has been proven by technical and economic estimations, one of the most important factors for successful promotion of reactor designs is their net efficiency, which must be not lower than 47%. A significant advantage of a reactor plant with the HTGR and gas-turbine power conversion unit over the steam cycle is considerable simplification of the power unit layout and reduction of the required equipment and systems (no steam generators, no turbine hall including steam lines, condenser, deaerator, etc.), which makes the gas-turbine power conversion unit more compact and less costly in production, operation and maintenance. However, in spite of this advantage, it seems that in the projects currently being developed, the potential of the gas-turbine cycle and high-temperature reactor to more efficiently generate electricity is not fully used. For example, in modern

  15. Conceptual study on HTGR-IS hydrogen supply system using organic hydrides

    International Nuclear Information System (INIS)

    Terada, Atsuhiko; Noguchi, Hiroki; Takegami, Hiroaki; Kamiji, Yu; Inagaki, Yoshiyuki

    2012-02-01

    We have proposed a hydrogen supply-chain system, which is a storage/supply system of large amount of hydrogen produced by HTGR-IS hydrogen production system. The organic chemical hydride method is one of the candidate techniques in the system for hydrogen storage and transportation. In this study, properties of organic hydrides and conventional hydrogen storage/supply system were surveyed to make use of the conceptual design of the hydrogen supply system using an organic hydrides method with VHTR-IS hydrogen production process (hydrogen production: 85,400 Nm 3 /h). Conceptual specifications of the main equipments were designed for the hydrogen supply system consisting of hydrogenation and dehydrogenation process. It was also clarified the problems of hydrogen supply system, such as energy efficiency and system optimization. (author)

  16. Comparison of three different collectors for process heat applications

    Science.gov (United States)

    Brunold, Stefan; Frey, R.; Frei, Ulrich

    1994-09-01

    In general vacuum tube collectors are used in solar process heat systems. Another possibility is to use transparent insulated flat plate collectors. A critical point however, is that most of the common transparent insulating materials can not withstand high temperatures because they consist of plastics. Thus, temperature resistive collector covers combining a high tranmisivity with a low U-value are required. One possibility is to use capillaries made of glass instead of plastics. Measurement results of collector efficiency and incident angle modifier will be presented as well as calculated energy gains for three different collectors: a vacuum tube collector (Giordano Ind., France), a CPC vacuum tube collector (microtherm Energietechnik Germany; a new flat plate collector using glass capillary as transparent insulation (SET, Germany).

  17. Industrial process heat case studies. [PROSYS/ECONMAT code

    Energy Technology Data Exchange (ETDEWEB)

    Hooker, D.W.; May, E.K.; West, R.E.

    1980-05-01

    Commercially available solar collectors have the potential to provide a large fraction of the energy consumed for industrial process heat (IPH). Detailed case studies of individual industrial plants are required in order to make an accurate assessment of the technical and economic feasibility of applications. This report documents the results of seven such case studies. The objectives of the case study program are to determine the near-term feasibility of solar IPH in selected industries, identify energy conservation measures, identify conditions of IPH systems that affect solar applications, test SERI's IPH analysis software (PROSYS/ECONOMAT), disseminate information to the industrial community, and provide inputs to the SERI research program. The detailed results from the case studies are presented. Although few near-term, economical solar applications were found, the conditions that would enhance the opportunities for solar IPH applications are identified.

  18. Environmental assessment for radioisotope heat source fuel processing and fabrication

    International Nuclear Information System (INIS)

    1991-07-01

    DOE has prepared an Environmental Assessment (EA) for radioisotope heat source fuel processing and fabrication involving existing facilities at the Savannah River Site (SRS) near Aiken, South Carolina and the Los Alamos National Laboratory (LANL) near Los Alamos, New Mexico. The proposed action is needed to provide Radioisotope Thermoelectric Generators (RTG) to support the National Aeronautics and Space Administration's (NASA) CRAF and Cassini Missions. Based on the analysis in the EA, DOE has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement is not required. 30 refs., 5 figs

  19. Process and apparatus for indirect-fired heating and drying

    Science.gov (United States)

    Abbasi, Hamid Ali; Chudnovsky, Yaroslav

    2005-04-12

    A method for heating flat or curved surfaces comprising injecting fuel and oxidant along the length, width or longitudinal side of a combustion space formed between two flat or curved plates, transferring heat from the combustion products via convection and radiation to the surface being heated on to the material being dried/heated, and recirculating at least 20% of the combustion products to the root of the flame.

  20. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-04-01

    In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure and a simultaneous loss of forced cool of the core

  1. Process of optimization of district heat production by utilizing waste energy from metallurgical processes

    Science.gov (United States)

    Konovšek, Damjan; Fužir, Miran; Slatinek, Matic; Šepul, Tanja; Plesnik, Kristijan; Lečnik, Samo

    2017-07-01

    In a consortium with SIJ (Slovenian Steel Group), Metal Ravne, the local community of Ravne na Koro\\vskem and the public research Institut Jožef Stefan, with its registered office in Slovenia, Petrol Energetika, d.o.o. set up a technical and technological platform of an innovative energy case for a transition of steel industry into circular economy with a complete energy solution called »Utilization of Waste Heat from Metallurgical Processes for District Heating of Ravne na Koro\\vskem. This is the first such project designed for a useful utilization of waste heat in steel industry which uses modern technology and innovative system solutions for an integration of a smart, efficient and sustainable heating and cooling system and which shows a growth potential. This will allow the industry and cities to make energy savings, to improve the quality of air and to increase the benefits for the society we live in. On the basis of circular economy, we designed a target-oriented co-operation of economy, local community and public research institute to produce new business models where end consumers are put into the centre. This innovation opens the door for steel industry and local community to a joint aim that is a transition into efficient low-carbon energy systems which are based on involvement of natural local conditions, renewable energy sources, the use of waste heat and with respect for the principles of sustainable development.

  2. Status of a reformer design for a modular HTGR in an in-line configuration

    International Nuclear Information System (INIS)

    Gluck, R.; Whitling, W.H.; Lipps, A.J.

    1984-01-01

    For the past several years the General Electric Company has had the technical lead on advanced concept studies for the Modular High Temperature Gas Cooled Reactor (HTGR) programs sponsored by the United States Department of Energy. The focus of the Modular Reactor System (MRS) effort is the development of a generic nuclear heat source capable of supplying heat to either a steam generator/electric cycle or a high temperature steam /methane reforming cycle. Some early ground rules for this study were that the reactor be designed for 950 deg. C direct cycle reforming and that the core be of the prismatic type. Since the prismatic core required control rods near the center of the core, the vertical in-line concept was selected to promote natural circulation cooling of the core for all potential transients except the depressurized core heatup transient. Although the requirement for a prismatic core has been eliminated for recent cost reduction studies, the vertical in-line configuration has been retained for its potential as the lowest cost configuration. This paper presents the results of recent design and analytical studies conducted to evaluate the feasibility of using a steam/methane reformer in a Vertical In-Line (VIL) arrangement with the generic nuclear heat source

  3. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  4. Auxiliary Heat Exchanger Flow Distribution Test

    International Nuclear Information System (INIS)

    Kaufman, J.S.; Bressler, M.M.

    1983-01-01

    The Auxiliary Heat Exchanger Flow Distribution Test was the first part of a test program to develop a water-cooled (tube-side), compact heat exchanger for removing heat from the circulating gas in a high-temperature gas-cooled reactor (HTGR). Measurements of velocity and pressure were made with various shell side inlet and outlet configurations. A flow configuration was developed which provides acceptable velocity distribution throughout the heat exchanger without adding excessive pressure drop

  5. Hydrogen production from coal using a nuclear heat source

    International Nuclear Information System (INIS)

    Quade, R.N.

    1977-01-01

    A strong candidate for hydrogen production in the intermediate time frame of 1990 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas - a feedstock which may not be available in large quantities in this time frame. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. Bench-scale experimental work on the hydrogasification of coal liquids is being carried out. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed and shows diminishing returns at process temperatures above about 1500 0 F. (author)

  6. Hydrogen production from coal using a nuclear heat source

    Science.gov (United States)

    Quade, R. N.

    1976-01-01

    A strong candidate for hydrogen production in the intermediate time frame of 1985 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed, and shows diminishing returns at process temperatures above about 1500 F. A possible scenario combining the relatively abundant and low-cost Western coal deposits with the Gulf Coast hydrogen users is presented which provides high-energy density transportation utilizing coal liquids and uranium.

  7. Modeling conductive heat transfer during high-pressure thawing processes: determination of latent heat as a function of pressure.

    Science.gov (United States)

    Denys, S; Van Loey, A M; Hendrickx, M E

    2000-01-01

    A numerical heat transfer model for predicting product temperature profiles during high-pressure thawing processes was recently proposed by the authors. In the present work, the predictive capacity of the model was considerably improved by taking into account the pressure dependence of the latent heat of the product that was used (Tylose). The effect of pressure on the latent heat of Tylose was experimentally determined by a series of freezing experiments conducted at different pressure levels. By combining a numerical heat transfer model for freezing processes with a least sum of squares optimization procedure, the corresponding latent heat at each pressure level was estimated, and the obtained pressure relation was incorporated in the original high-pressure thawing model. Excellent agreement with the experimental temperature profiles for both high-pressure freezing and thawing was observed.

  8. SOLTECH 92 proceedings: Solar Process Heat Program. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1992-03-01

    This document is a limited Proceedings, documenting the presentations given at the symposia conducted by the US Department of Energy`s (DOE) Solar Industrial Program and Solar Thermal Electrical Program at SOLTECH92. The SOLTECH92 national solar energy conference was held in Albuquerque, New Mexico during the period February 17--20, 1992. The National Renewable Energy Laboratory manages the Solar Industrial Program; Sandia National Laboratories (Albuquerque) manages the Solar Thermal Electric Program. The symposia sessions were as follows: (1) Solar Industrial Program and Solar Thermal Electric Program Overviews, (2) Solar Process Heat Applications, (3) Solar Decontamination of Water and Soil; (4) Solar Building Technologies, (5) Solar Thermal Electric Systems, (6) PV Applications and Technologies. For each presentation given in these symposia, these Proceedings provide a one- to two-page abstract and copies of the viewgraphs and/or 35mm slides utilized by the speaker. Some speakers provided additional materials in the interest of completeness. The materials presented in this document were not subjected to a peer review process.

  9. Market development directory for solar industrial process heat systems

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-02-01

    The purpose of this directory is to provide a basis for market development activities through a location listing of key trade associations, trade periodicals, and key firms for three target groups. Potential industrial users and potential IPH system designers were identified as the prime targets for market development activities. The bulk of the directory is a listing of these two groups. The third group, solar IPH equipment manufacturers, was included to provide an information source for potential industrial users and potential IPH system designers. Trade associates and their publications are listed for selected four-digit Standard Industrial Code (SIC) industries. Since industries requiring relatively lower temperature process heat probably will comprise most of the near-term market for solar IPH systems, the 80 SIC's included in this chapter have process temperature requirements less than 350/sup 0/F. Some key statistics and a location list of the largest plants (according to number of employees) in each state are included for 15 of the 80 SIC's. Architectural/engineering and consulting firms are listed which are known to have solar experience. Professional associated and periodicals to which information on solar IPH sytstems may be directed also are included. Solar equipment manufacturers and their associations are listed. The listing is based on the SERI Solar Energy Information Data Base (SEIDB).

  10. SOLTECH 1992 proceedings: Solar Process Heat Program, volume 1

    Science.gov (United States)

    1992-03-01

    This document is a limited Proceedings, documenting the presentations given at the symposia conducted by the U.S. Department of Energy's (DOE) Solar Industrial Program and Solar Thermal Electrical Program at SOLTECH92. The SOLTECH92 national solar energy conference was held in Albuquerque, New Mexico during the period February 17-20, 1992. The National Renewable Energy Laboratory manages the Solar Industrial Program; Sandia National Laboratories (Albuquerque) manages the Solar Thermal Electric Program. The symposia sessions were as follows: (1) Solar Industrial Program and Solar Thermal Electric Program Overviews, (2) Solar Process Heat Applications, (3) Solar Decontamination of Water and Soil, (4) Solar Building Technologies, (5) Solar Thermal Electric Systems, and (6) Photovoltaic (PV) Applications and Technologies. For each presentation given in these symposia, these Proceedings provide a one- to two-page abstract and copies of the viewgraphs and/or 35 mm slides utilized by the speaker. Some speakers provided additional materials in the interest of completeness. The materials presented in this document were not subjected to a peer review process.

  11. Candidate thermal energy storage technologies for solar industrial process heat applications

    Science.gov (United States)

    Furman, E. R.

    1979-01-01

    A number of candidate thermal energy storage system elements were identified as having the potential for the successful application of solar industrial process heat. These elements which include storage media, containment and heat exchange are shown.

  12. 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)

  13. Fluidized combustion of beds of large, dense particles in reprocessing HTGR fuel

    International Nuclear Information System (INIS)

    Young, D.T.

    1977-03-01

    Fluidized bed combustion of graphite fuel elements and carbon external to fuel particles is required in reprocessing high-temperature gas-cooled reactor (HTGR) cores for recovery of uranium. This burning process requires combustion of beds containing both large particles and very dense particles as well as combustion of fine graphite particles which elutriate from the bed. Equipment must be designed for optimum simplicity and reliability as ultimate operation will occur in a limited access ''hot cell'' environment. Results reported in this paper indicate that successful long-term operation of fuel element burning with complete combustion of all graphite fines leading to a fuel particle product containing <1% external carbon can be performed on equipment developed in this program

  14. Containerless Heating Process of a Deeply Undercooled Metal Droplet by Electrostatic Levitation

    International Nuclear Information System (INIS)

    Wang Fei-Long; Dai Bin; Liu Xue-Feng; Sun Yi-Ning; Sun Zhi-Bin; Yu Qiang; Zhai Guang-Jie

    2015-01-01

    We present the containerless heating process of a deeply undercooled metal droplet by electrostatic levitation. The problem of surface charge loss in the heating process is discussed and specific formulas are given to describe the basic process of charge supplement by the photoelectric and thermoelectric effects. The pure metal zirconium is used to be melted and solidified to analyze the heating process. The temperature-time curve clearly shows the features including melting, undercooling, recalescence and solid-state phase transformation. (paper)

  15. Present activity of the feasibility study of HTGR-GT system

    International Nuclear Information System (INIS)

    Muto, Y.; Miyamoto, Y.; Shiozawa, S.

    2001-01-01

    In JAERI a feasibility study of the High Temperature Gas-cooled Reactor-Gas Turbine (HTGR-GT) system has been carried out since January, 1997 as an assigned work by the Science and Technology Agency. The study aims at obtaining a promising concept of HTGR-GT system that yields a high thermal efficiency and at the same time is economically competitive. Designs of a few candidate systems will be undertaken and their power generation costs will be evaluated in parallel with design works, some experimental works such as the fabrication of a plate-fin type heat exchanger core and material tests will be carried out. The study will be continued till 2000 fiscal year. In 1997 fiscal year, a preliminary design of a direct cycle plant of 600 MWt was developed. A reactor inlet gas temperature of 460 deg. C, a reactor outlet gas temperature of 850 deg. C and a helium gas pressure of 6MPa were selected. Some advanced technologies were adopted such as a monolithic fuel compact and a control rod sheath made of carbon/carbon composite material. They were very effective to enhance the heat transfer of fuel and to reduce the core bypass flow. As a result, a power density of 6MW/m 3 and the maximum burnup of 10 5 MWD/ton were achieved. A single-shaft horizontal turbomachine of 3600 rpm was selected to ease the mechanical design of the rotor supported by magnetic bearings. The turbine, two compressors, a generator and six units of intercooler were placed in a turbine vessel, Plate-fin type recuperator and precooler are installed in a vertical heat exchanger vessel. By this design, a net thermal efficiency of 45.7% is expected to be achieved. To develop a high performance plate-fin recuperator, a core model of W200 mm x L200 mm x H200 mm with small fin size of 1.15 mm height was fabricated and as a result of tests, leak tightness, component strength and bonding appearance were found to be satisfactory. In 1998 fiscal year, a design of a direct cycle plant of 300 MWt is undertaken. The

  16. Recent evolution of HTGR instrumentation in the USA

    International Nuclear Information System (INIS)

    Rodriguez, C.

    1982-06-01

    The reactor instrumentation system for the 2240 MW(t) HTGR includes ex-core neutron detectors for automatic nuclear power control, separate ex-core neutron detectors for automatic protection purposes (reactor trip), reactor core outlet thermocouples that measure the temperature of the primary coolant (helium) as it exits the nuclear core, cold helium thermocouples that measure the temperature of the primary coolant as it enters the core, external pressure differential gages that measure primary coolant flow, in-core fission chambers that are utilized to map neutron flux, and ex-core primary coolant moisture monitors. All of these subsystems, except for the in-core flux mapping units, are also part of the Fort St. Vrain HTGR, which has provided significant experience for the design of the new system. In-core flux mapping is not necessary at FSV for normal operation because its relatively small core is fairly ''visible'' from the location of the ex-core instruments. However, temporary in-core fission couples, microphones, and displacement sensors, as well as sensitive ex-core accelerometers were utilized to identify periodic core block lateral movement and measure neutron flux and primary coolant temperatures. A search for in-core sensors to facilitate mapping neutron flux distributions in the larger core of the 2240 MW(t) HTGR has led to the selection of a high temperature fission chamber, which has been tested up to 1000 deg. C at General Atomic. The chamber shows adequate signal to noise ratio and repeatability. Other reactor instruments planned for the 2240 MW(t) are of the FSV type (i.e. thermocouples) or improved versions of the FSV design (i.e. moisture monitors). New concepts such as acoustic thermometers are also being considered

  17. HTGR technology development: status and direction

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1982-01-01

    During the last two years there has been an extensive and comprehensive effort expended primarily by General Atomic (GA) in generating a revised technology development plan. Oak Ridge National Laboratory (ORNL) has assisted in this effort, primarily through its interactions over the past years in working together with GA in technology development, but also through detailed review of the initial versions of the technology development plan as prepared by GA. The plan covers Fuel Technology, Materials Technology (including metals, graphite, and ceramics), Plant Technology (including methods, safety, structures, systems, heat exchangers, control and electrical, and mechanical), and Component Design Verification and Support areas

  18. The market for HTGR type reactors

    International Nuclear Information System (INIS)

    Roehler, E.

    1986-01-01

    High-temperature-reactors with pebble-bed-reactor cores as a progressive reactor line, have been developed by BBC/HRB the Federal Republic of Germany over a period of 27 years and will soon be mature to be introduced to the market. They represent an important innovation in the field of reactor engineering. Due to its high degree of applicability on the power and heat market and its high flexibility regarding the site and fuel cycle the HTR is extremely suitable for providing energy to consumers, especially in countries using nuclear energy supply for the first time. (orig.) [de

  19. Waste heat and water recovery opportunities in California tomato paste processing

    International Nuclear Information System (INIS)

    Amón, Ricardo; Maulhardt, Mike; Wong, Tony; Kazama, Don; Simmons, Christopher W.

    2015-01-01

    Water and energy efficiency are important for the vitality of the food processing industry as demand for these limited resources continues to increase. Tomato processing, which is dominated by paste production, is a major industry in California – where the majority of tomatoes are processed in the United States. Paste processing generates large amounts of condensate as moisture is removed from the fruit. Recovery of the waste heat in this condensate and reuse of the water may provide avenues to decrease net energy and water use at processing facilities. However, new processing methods are needed to create demand for the condensate waste heat. In this study, the potential to recover condensate waste heat and apply it to the tomato enzyme thermal inactivation processing step (the hot break) is assessed as a novel application. A modeling framework is established to predict heat transfer to tomatoes during the hot break. Heat recovery and reuse of the condensate water are related to energy and monetary savings gained through decreased use of steam, groundwater pumping, cooling towers, and wastewater processing. This analysis is informed by water and energy usage data from relevant unit operations at a commercial paste production facility. The case study indicates potential facility seasonal energy and monetary savings of 7.3 GWh and $166,000, respectively, with most savings gained through reduced natural gas use. The sensitivity of heat recovery to various process variables associated with heat exchanger design and processing conditions is presented to identify factors that affect waste heat recovery. - Highlights: • The potential to recovery waste heat in tomato paste processing is examined. • Heat transfer from evaporator condensate to tomatoes in the hot break is modeled. • Processing facility data is used in model to predict heat recovery energy savings. • The primary benefit of heat recovery is reduced use of natural gas in boilers. • Reusing

  20. Project summary plan for HTGR recycle reference facility

    International Nuclear Information System (INIS)

    Baxter, B.J.

    1979-11-01

    A summary plan is introduced for completing conceptual definition of an HTGR Recycle Reference Facility (HRRF). The plan describes a generic project management concept, often referred to as the requirements approach to systems engineering. The plan begins with reference flow sheets and provides for the progressive evolution of HRRF requirements and definition through feasibility, preconceptual, and conceptual phases. The plan lays end-to-end all the important activities and elements to be treated during each phase of design. Identified activities and elements are further supported by technical guideline documents, which describe methodology, needed terminology, and where relevant a worked example

  1. Recent developments in graphite. [Use in HTGR and aerospace

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.

  2. A reactivity accidents simulation of the Fort Saint Vrain HTGR

    International Nuclear Information System (INIS)

    Fainer, Gerson

    1980-01-01

    A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

  3. Automatic particle-size analysis of HTGR recycle fuel

    International Nuclear Information System (INIS)

    Mack, J.E.; Pechin, W.H.

    1977-09-01

    An automatic particle-size analyzer was designed, fabricated, tested, and put into operation measuring and counting HTGR recycle fuel particles. The particle-size analyzer can be used for particles in all stages of fabrication, from the loaded, uncarbonized weak acid resin up to fully-coated Biso or Triso particles. The device handles microspheres in the range of 300 to 1000 μm at rates up to 2000 per minute, measuring the diameter of each particle to determine the size distribution of the sample, and simultaneously determining the total number of particles. 10 figures

  4. Treatment of operator actions in the HTGR risk assessment study

    International Nuclear Information System (INIS)

    Fleming, K.N.; Silady, F.A.; Hannaman, G.W.

    1979-12-01

    Methods are presented for the treatment of operator actions, developed in the AIPA risk assessment study. Some examples are given of how these methods were applied to the analysis of potential HTGR accidents. Realistic predictions of accident risks required a balanced treatment of both beneficial and detrimental actions and responses of human operators and maintenance crews. Th essential elements of the human factors methodology used in the AIPA study include event tree and fault tree analysis, time-dependent operator response and repair models, a method for quantifying common cause failure probabilities, and synthesis of relevant experience data for use in these models

  5. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  6. The calculation - experimental investigations of the HTGR fuel element construction

    International Nuclear Information System (INIS)

    Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

    1985-01-01

    One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

  7. Prediction of heat treatment in food processing machinery

    DEFF Research Database (Denmark)

    Karlson, Torben; Friis, Alan; Szabo, Peter

    1997-01-01

    The velocity and temperature fields of a shear thinning fluid in a co-rotating disc scraped surface heat exchanger (CDHE) are calculated using the finite element method. By tracking and timingparticles through the heat exchanger residence time and thermal time distributions are computed. The resi...

  8. Research on Heat Exchange Process in Aircraft Air Conditioning System

    Science.gov (United States)

    Chichindaev, A. V.

    2017-11-01

    Using of heat-exchanger-condenser in the air conditioning system of the airplane Tu-204 (Boeing, Airbus, Superjet 100, MS-21, etc.) for cooling the compressed air by the cold air with negative temperature exiting the turbine results in a number of operational problems. Mainly it’s frosting of the heat exchange surface, which is the cause of live-section channels frosting, resistance increasing and airflow in the system decreasing. The purpose of this work is to analyse the known freeze-up-fighting methods for heat-exchanger-condenser, description of the features of anti-icing protection and offering solutions to this problem. For the problem of optimizing the design of heat exchangers in this work used generalized criterion that describes the ratio of thermal resistances of cold and hot sections, which include: the ratio of the initial values of heat transfer agents flow state; heat exchange surface finning coefficients; factors which describes the ratio of operating parameters and finning area. By controlling the ratio of the thermal resistances can be obtained the desired temperature of the heat exchange surface, which would prevent freezing. The work presents the results of a numerical study of the effect of different combinations of regime and geometrical factors changes on reduction of the heat-exchanger-condenser freezing surface area, including using of variable ratio of thermal resistances.

  9. Correlation of heat transfer coefficient in quenching process using ABAQUS

    Science.gov (United States)

    Davare, Sandeep Kedarnath; Balachandran, G.; Singh, R. K. P.

    2018-04-01

    During the heat treatment by quenching in a liquid medium the convective heat transfer coefficient plays a crucial role in the extraction of heat. The heat extraction ultimately influences the cooling rate and hence the hardness and mechanical properties. A Finite Element analysis of quenching a simple flat copper sample with different orientation of sample and with different quenchant temperatures were carried out to check and verify the results obtained from the experiments. The heat transfer coefficient (HTC) was calculated from temperature history in a simple flat copper disc sample experimentally. This HTC data was further used as input to simulation software and the cooling curves were back calculated. The results obtained from software and using experimentation shows nearly consistent values.

  10. TRANTHAC-1: transient thermal-hydraulic analysis code for HTGR core of multi-channel model

    International Nuclear Information System (INIS)

    Sato, Sadao; Miyamoto, Yoshiaki

    1980-08-01

    The computer program TRANTHAC-1 is for predicting thermal-hydraulic transient behavior in HTGR's core of pin-in-block type fuel elements, taking into consideration of the core flow distribution. The program treats a multi-channel model, each single channel representing the respective column composed of fuel elements. The fuel columns are grouped in flow control regions; each region is provided with an orifice assembly. In the region, all channels are of the same shape except one channel. Core heat is removed by downward flow of the control through the channel. In any transients, for given time-dependent power, total core flow, inlet coolant temperature and coolant pressure, the thermal response of the core can be determined. In the respective channels, the heat conduction in radial and axial direction are represented. And the temperature distribution in each channel with the components is calculated. The model and usage of the program are described. The program is written in FORTRAN-IV for computer FACOM 230-75 and it is composed of about 4,000 cards. The required core memory is about 75 kilowords. (author)

  11. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-01-01

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry, and the utilities. Near-term development is focused on electricity generation. The top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. Quantitatively, this requires that the design meet the US Environmental Protection Agency's Protective Action Guides at the site boundary and hence preclude the need for sheltering or evacuation of the public. To meet these stringent safety requirements and at the same time provide a cost competitive design requires the innovative use of the basic high temperature gas-cooled reactor features of ceramic fuel, helium coolant, and a graphite moderator. The specific fuel composition and core size and configuration have been selected to the use the natural characteristics of these materials to develop significantly higher margins of safety. In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure of a simultaneous loss of forced cooling of the core. The challenges to control of heat generation consider not only the failure to insert the reactivity control systems but also the withdrawal of control rods. Finally, challenges to control of chemical attack of the ceramic-coated fuel are considered, including catastrophic failure of the steam generator, which allows water ingress, or failure of the pressure vessels, which allows air ingress. The plant's response to these extreme challenges is not dependent on operator action, and the events considered encompass conceivable operator errors

  12. Developing maintenance technologies for FBR's heat exchanger units by advanced laser processing

    International Nuclear Information System (INIS)

    Nishimura, Akihiko; Shimada, Yukihiro

    2011-01-01

    Laser processing technologies were developed for the purpose of maintenance of FBR's heat exchanger units. Ultrashort laser processing fabricated fiber Bragg grating sensor for seismic monitoring. Fiber laser welding with a newly developed robot system repair cracks on inner wall of heat exchanger tubes. Safety operation of the heat exchanger units will be improved by the advanced laser processing technologies. These technologies are expected to be applied to the maintenance for the next generation FBRs. (author)

  13. LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR

    Directory of Open Access Journals (Sweden)

    S Simbolon

    2015-03-01

    Full Text Available LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR. The The aim of this research is to produce thousands of microsphere ADU (Ammonium Diuranate gels by using external gelation for an intermediate HTGR (High Temperature Gas-cooled Reactor nuclear fuel in laboratory-scale. Microsphere ADU gels were based on sol-solution which was made from a homogeneous mixture of ADUN (Acid Deficient Uranyl Nitrate which was containing uranyl ion in high concentration, a water soluble organic compound PVA (Polyvinyl Alcohol and THFA (Tetrahydrofurfuryl Alcohol. The simple unified home made laboratory experimental machine was developed to replace test tube experiment method which was once used due to a tiny amount of microsphere ADU gels produced. It consists of four main parts: tank filled sol-solution connecting to peristaltic pump and vibrating nozzle, preliminary gelation and gelation column. The machine has successfully converted 150 mL sol-solution into thousands of drops which produced 120 - 130 drops in each minute in steady state in ammonia gas free sector. Preliminary gelation reaction was carried out in ammonia gas sector where drops react with ammonia gas in a bat an eye followed by gelation reaction in column containing ammonia solution 7 M. In ageing process, ADU gels were collected and submerged into a vessel containing ammonia solution which was shaken for 1 hour in a shaker device. Isopropyl alcohol (90% solution was used to wash ADU gels and a digital camera was used to measured spherical form of ADU gels. Diameters in spherical spheroid form were found between 1.8 mm until 2.2 mm. The spherical purity of ADU gels were 10% - 20% others were oblate, prolate spheroid and microsphere which have hugetiny of dimples on the surface.   PRODUKSI GEL ADU SKALA LABORATORIUM DENGAN MENGGUNAKAN GELASI EKSTERNAL UNTUK BAHAN BAKAR ANTARA HTGR. Penelitian ini bertujuan untuk membuat ribuan gel bulat ADU (Ammonium

  14. Heat recovery networks synthesis of large-scale industrial sites: Heat load distribution problem with virtual process subsystems

    International Nuclear Information System (INIS)

    Pouransari, Nasibeh; Maréchal, Francois

    2015-01-01

    Highlights: • Synthesizing industrial size heat recovery network with match reduction approach. • Targeting TSI with minimum exchange between process subsystems. • Generating a feasible close-to-optimum network. • Reducing tremendously the HLD computational time and complexity. • Generating realistic network with respect to the plant layout. - Abstract: This paper presents a targeting strategy to design a heat recovery network for an industrial plant by dividing the system into subsystems while considering the heat transfer opportunities between them. The methodology is based on a sequential approach. The heat recovery opportunity between process units and the optimal flow rates of utilities are first identified using a Mixed Integer Linear Programming (MILP) model. The site is then divided into a number of subsystems where the overall interaction is resumed by a pair of virtual hot and cold stream per subsystem which is reconstructed by solving the heat cascade inside each subsystem. The Heat Load Distribution (HLD) problem is then solved between those packed subsystems in a sequential procedure where each time one of the subsystems is unpacked by switching from the virtual stream pair back into the original ones. The main advantages are to minimize the number of connections between process subsystems, to alleviate the computational complexity of the HLD problem and to generate a feasible network which is compatible with the minimum energy consumption objective. The application of the proposed methodology is illustrated through a number of case studies, discussed and compared with the relevant results from the literature

  15. CONTEMPT-G computer program and its application to HTGR containments

    International Nuclear Information System (INIS)

    Macnab, D.I.

    1976-03-01

    The CONTEMPT-G computer program has been developed by General Atomic Company to simulate the temperature-pressure response of a containment atmosphere to postulated depressurization of High-Temperature Gas-Cooled Reactor (HTGR) primary or secondary coolant circuits. The mathematical models currently used in the code are described, and applications of the code in examples of the atmospheric response of a representative containment to a variety of postulated HTGR accident conditions are presented. In particular, maximum containment temperature and pressure, equilibrated long-term prestressed concrete reactor vessel and containment pressures, and peak containment conditions following steam pipe ruptures are examined for a representative 770-MW(e) HTGR

  16. Dynamics and control modeling of the closed-cycle gas turbine (GT-HTGR) power plant

    International Nuclear Information System (INIS)

    Bardia, A.

    1980-02-01

    The simulation if presented for the 800-MW(e) two-loop GT-HTGR plant design with the REALY2 transient analysis computer code, and the modeling of control strategies called for by the inherently unique operational requirements of a multiple loop GT-HTGR is described. Plant control of the GT-HTGR is constrained by the nature of its power conversion loops (PCLs) in which the core cooling flow and the turbine flow are directly related and thus changes in flow affect core cooling as well as turbine power. Additionally, the high thermal inertia of the reactor core precludes rapid changes in the temperature of the turbine inlet flow

  17. Effects of the HTGR-gas turbine on national reactor strategies

    International Nuclear Information System (INIS)

    Ligon, D.M.; Brogli, R.H.

    1979-11-01

    A specific role for the HTGR in a national energy strategy is examined. The issue is addressed in two ways. First, the role of the HTGR-GT Binary cycle plant is examined in a national energy strategy based on symbiosis between fast breeder and advanced converter reactors utilizing the thorium U233 fuel cycle. Second, the advantages of the HTGR-GT dry-cooled plant operating in arid regions is examined and compared with a dry-cooled LWR. An event tree analysis of potential benefits is applied

  18. In situ conversion process utilizing a closed loop heating system

    Science.gov (United States)

    Sandberg, Chester Ledlie [Palo Alto, CA; Fowler, Thomas David [Houston, TX; Vinegar, Harold J [Bellaire, TX; Schoeber, Willen Jan Antoon Henri

    2009-08-18

    An in situ conversion system for producing hydrocarbons from a subsurface formation is described. The system includes a plurality of u-shaped wellbores in the formation. Piping is positioned in at least two of the u-shaped wellbores. A fluid circulation system is coupled to the piping. The fluid circulation system is configured to circulate hot heat transfer fluid through at least a portion of the piping to form at least one heated portion of the formation. An electrical power supply is configured to provide electrical current to at least a portion of the piping located below an overburden in the formation to resistively heat at least a portion of the piping. Heat transfers from the piping to the formation.

  19. Modelling of heat and mass transfer processes in neonatology

    Energy Technology Data Exchange (ETDEWEB)

    Ginalski, Maciej K [FLUENT Europe, Sheffield Business Park, Europa Link, Sheffield S9 1XU (United Kingdom); Nowak, Andrzej J [Institute of Thermal Technology, Silesian University of Technology, Konarskiego 22, 44-100 Gliwice (Poland); Wrobel, Luiz C [School of Engineering and Design, Brunel University, Uxbridge UB8 3PH (United Kingdom)], E-mail: maciej.ginalski@ansys.com, E-mail: Andrzej.J.Nowak@polsl.pl, E-mail: luiz.wrobel@brunel.ac.uk

    2008-09-01

    This paper reviews some of our recent applications of computational fluid dynamics (CFD) to model heat and mass transfer problems in neonatology and investigates the major heat and mass transfer mechanisms taking place in medical devices such as incubators and oxygen hoods. This includes novel mathematical developments giving rise to a supplementary model, entitled infant heat balance module, which has been fully integrated with the CFD solver and its graphical interface. The numerical simulations are validated through comparison tests with experimental results from the medical literature. It is shown that CFD simulations are very flexible tools that can take into account all modes of heat transfer in assisting neonatal care and the improved design of medical devices.

  20. Modelling of heat and mass transfer processes in neonatology

    International Nuclear Information System (INIS)

    Ginalski, Maciej K; Nowak, Andrzej J; Wrobel, Luiz C

    2008-01-01

    This paper reviews some of our recent applications of computational fluid dynamics (CFD) to model heat and mass transfer problems in neonatology and investigates the major heat and mass transfer mechanisms taking place in medical devices such as incubators and oxygen hoods. This includes novel mathematical developments giving rise to a supplementary model, entitled infant heat balance module, which has been fully integrated with the CFD solver and its graphical interface. The numerical simulations are validated through comparison tests with experimental results from the medical literature. It is shown that CFD simulations are very flexible tools that can take into account all modes of heat transfer in assisting neonatal care and the improved design of medical devices

  1. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  2. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  3. Tribological study on machine elements of HTGR components

    International Nuclear Information System (INIS)

    Nemoto, M.; Asanabe, S.; Kawaguchi, K.; Ono, S.; Oyamada, T.

    1980-01-01

    There are some tribological features peculiar to machines used in a high-temperature gas-cooled reactor (HTGR) plant. In this kind of plant, water-lubricated bearing combined with the buffer gas sealing system and/or gas-lubricated bearings are often applied in order to prevent degrading of the purity of coolant helium gas. And, it is essential for the reliability and safety design of the sliding members in the HTGR to obtain fundamental data on their friction and wear in high-temperature helium atmosphere. In this paper, the results of tests on these bearings and sliding members are introduced, which are summarized as follows: (1) Water-lubricated shrouded step thrust bearing and buffer gas sealing system were tested separately under the conditions simulated to those of circulators used in commercial plants. The results showed that each elements satisfies the requirements. (2) A hydrostatically gas-lubricated, pivoted pad journal bearing with a moat-shaped rectangular groove is found to be promising for use as a high-load bearing, which is indispensable for the development of a large-type circulator. (3) Use of ceramic coating and carbon graphite materials is effective for the prevention of adhesive wear which is apt to occur in metal-to-metal combinations. (author)

  4. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

  5. Tribological study on machine elements of HTGR components

    International Nuclear Information System (INIS)

    Nemoto, Masaaki; Ono, Shigeharu; Asanabe, Sadao; Kawaguchi, Katsuyuki; Oyamada, Tetsuya.

    1981-11-01

    There are some tribological features peculiar to machines used in a high-temperature gas-cooled reactor (HTGR) plant. In this kind of plant, water-lubricated bearing combined with the buffer gas sealing system and/or gas-lubricated bearings are often applied in order to prevent degrading of the purity of coolant helium gas. And, it is essential for the reliability and safety design of the sliding members in the HTGR to obtain fundamental data on their friction and wear in high-temperature helium atmosphere. In this paper, the results of tests on these bearings and sliding members are introduced, which are summarized as follows: (1) Water-lubricated shrouded step thrust bearing and buffer gas sealing system were tested separately under the condition simulated to those of circulators used in commercial plants. The results showed that each elements satisfies the requirements. (2) A hydrostatically gas-lubricated, pivoted pad journal bearing with a moat-shaped rectangular groove is found to be promising for use as a high-load bearing, which is indispensable for the development of a large-type circulator. (3) Use of ceramic coating and carbon graphite materials is effective for the prevention of adhesive wear which is apt to occur in metal-to-metal combinations. (author)

  6. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

    In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

  7. Process for extracting residual heat and device for the ultimate absorption of heat for nuclear reactors

    International Nuclear Information System (INIS)

    Bernard, Lawrence Jr.

    1980-01-01

    This invention concerns a 'heat sink' or device for the ultimate absorption of heat for electric power stations using the most widespread thermal neutron nuclear reactors, namely 'light water' reactors such as boiling or pressurized water reactors. The residual heat given off by these reactors can be safely extracted with this method by using dry cooling. However, the invention does not concern the problems arising from the cooling of the steam used for actuating the steam turbine nor the cooling of the steam exhausted by the turbine or coming from it, but it does concern the 'safety' part of the nuclear power station in which the residual heat discharged in the reactor is controlled and dissipated [fr

  8. Heat and work integration: Fundamental insights and applications to carbon dioxide capture processes

    International Nuclear Information System (INIS)

    Fu, Chao; Gundersen, Truls

    2016-01-01

    Highlights: • The problem definition of heat and work integration is introduced. • The fundamental insights of heat and work integration are presented. • The design methodology is illustrated with two small test examples. • Applications of to three carbon dioxide capture processes are presented. - Abstract: The integration of heat has achieved a notable success in the past decades. Pinch Analysis is a well-established methodology for heat integration. Work is an equally important thermodynamic parameter. The enthalpy of a process stream can be changed by the transfer of heat and/or work. Heat and work are actually interchangeable and can thus be integrated. For example, compression processes consume more work at higher temperatures, however, the compression heat may be upgraded and utilized; expansion processes produce more work at higher temperatures, however, more heat may be required. The classical heat integration problem is thus extended to a new research topic about the integration of both heat and work. The aim of this paper is to present the problem definition, fundamental thermodynamic insights and industrial applications of heat and work integration. The results from studies on the three carbon dioxide capture processes show that significant energy savings can be achieved by proper heat and work integration. In the oxy-combustion process, the work consumption for cryogenic air separation is reduced by 10.1%. In the post-combustion membrane separation process, the specific work consumption for carbon dioxide separation is reduced by 12.9%. In the membrane air separation process, the net work consumption (excluding heat consumption) is reduced by 90%.

  9. 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

  10. Production control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.

    1979-06-01

    Purpose of this report was (1) to determine which techniques (Kalman Filter, weighted least squares, Shewhart control chart) are capable of detecting drift or step changes earliest in a manufacturing process, and (2) what method would work well in maintaining the manufacturing process at an acceptable level of quality. To solve part (1) simulation studies were performed for various test cases of interest. No single technique was superior in all of these cases, but the Kalman Filter appeared to be more robust to various process changes. The weighted least squares did a good job when the weight was near unity (0.9977) and failed when the weight was small (0.63). The Shewhart control chart is better for detecting step changes than for trends. Several methods wre compared to try to answer part (2). In this report the model building and forecasting was done using the methods of Box and Jenkins

  11. Measurement of heat pump processes induced by laser radiation

    Science.gov (United States)

    Garbuny, M.; Henningsen, T.

    1983-01-01

    A series of experiments was performed in which a suitably tuned CO2 laser, frequency doubled by a Tl3AsSe37 crystal, was brought into resonance with a P-line or two R-lines in the fundamental vibration spectrum of CO. Cooling or heating produced by absorption in CO was measured in a gas-thermometer arrangement. P-line cooling and R-line heating could be demonstrated, measured, and compared. The experiments were continued with CO mixed with N2 added in partial pressures from 9 to 200 Torr. It was found that an efficient collisional resonance energy transfer from CO to N2 existed which increased the cooling effects by one to two orders of magnitude over those in pure CO. Temperature reductions in the order of tens of degrees Kelvin were obtained by a single pulse in the core of the irradiated volume. These measurements followed predicted values rather closely, and it is expected that increase of pulse energies and durations will enhance the heat pump effects. The experiments confirm the feasibility of quasi-isentropic engines which convert laser power into work without the need for heat rejection. Of more immediate potential interest is the possibility of remotely powered heat pumps for cryogenic use, such applications are discussed to the extent possible at the present stage.

  12. Selected studies in HTGR reprocessing development

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-03-01

    Recent work at ORNL on hot cell studies, off-gas cleanup, and waste handling is reviewed. The work includes small-scale burning tests with irradiated fuels to study fission product release, development of the KALC process for the removal of 85 Kr from a CO 2 stream, preliminary work on a nonfluidized bed burner, solvent extraction studies including computer modeling, characterization of reprocessing wastes, and initiation of a development program for the fixation of 14 C as CaCO 3

  13. Systematic approach to optimal design of induction heating installations for aluminum extrusion process

    Science.gov (United States)

    Zimin, L. S.; Sorokin, A. G.; Egiazaryan, A. S.; Filimonova, O. V.

    2018-03-01

    An induction heating system has a number of inherent benefits compared to traditional heating systems due to a non-contact heating process. It is widely used in vehicle manufacture, cast-rolling, forging, preheating before rolling, heat treatment, galvanizing and so on. Compared to other heating technologies, induction heating has the advantages of high efficiency, fast heating rate and easy control. The paper presents a new systematic approach to the design and operation of induction heating installations (IHI) in aluminum alloys production. The heating temperature in industrial complexes “induction heating - deformation” is not fixed in advance, but is determined in accordance with the maximization or minimization of the total economic performance during the process of metal heating and deformation. It is indicated that the energy efficient technological complex “IHI – Metal Forming (MF)” can be designed only with regard to its power supply system (PSS). So the task of designing systems of induction heating is to provide, together with the power supply system and forming equipment, the minimum energy costs for the metal retreating.

  14. Integrated design and optimization of technologies for utilizing low grade heat in process industries

    International Nuclear Information System (INIS)

    Kwak, Dong-Hun; Binns, Michael; Kim, Jin-Kuk

    2014-01-01

    Highlights: • Implementation of a modeling and design framework for the utilization of low grade heat. • Application of process simulator and optimization techniques for the design of technologies for heat recovery. • Systematic and holistic exploitation for the recovery of industrial low grade heat. • Demonstration of the applicability and benefit of integrated design and optimization framework through a case study. - Abstract: The utilization of low grade heat in process industries has significant potential for improving site-wide energy efficiency. This paper focuses on the techno-economic analysis of key technologies for energy recovery and re-use, namely: Organic Rankine Cycles (ORC), boiler feed water heating, heat pumping and absorption refrigeration in the context of process integration. Process modeling and optimization in a holistic manner identifies the optimal integrated configuration of these technologies, with rigorous assessment of costs and technical feasibility of these technologies. For the systematic screening and evaluation of design options, detailed process simulator models are evaluated and optimization proceeds subject to design constraints for the particular economic scenarios where technology using low grade heat is introduced into the process site. Case studies are presented to illustrate how the proposed modeling and optimization framework can be useful and effective in practice, in terms of providing design guidelines and conceptual insights for the application of technologies using low grade heat. From the case study, the best options during winter are the ORC giving a 6.4% cost reduction for the ideal case with low grade heat available at a fixed temperature and boiler feed water heating giving a 2.5% cost reduction for the realistic case with low grade heat available at a range of temperatures. Similarly during summer boiler feed water heating was found to be the best option giving a 3.1% reduction of costs considering a

  15. Computational simulation of laser heat processing of materials

    Science.gov (United States)

    Shankar, Vijaya; Gnanamuthu, Daniel

    1987-04-01

    A computational model simulating the laser heat treatment of AISI 4140 steel plates with a CW CO2 laser beam has been developed on the basis of the three-dimensional, time-dependent heat equation (subject to the appropriate boundary conditions). The solution method is based on Newton iteration applied to a triple-approximate factorized form of the equation. The method is implicit and time-accurate; the maintenance of time-accuracy in the numerical formulation is noted to be critical for the simulation of finite length workpieces with a finite laser beam dwell time.

  16. Granular effect on the effective cross sections in the HTGR type reactors

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de.

    1975-01-01

    Effective cross section of bars for HTGR is studied from the point of view of heterogeneity. Microscopical heterogeneity due to grains is represented by a self-shielding factor, which is well determined [pt

  17. 2000 MW(t) HTGR-DC-GT Modesto Site dry cooled model 346 concice

    International Nuclear Information System (INIS)

    1979-07-01

    Construction information is presented for a 800 MW(e) HTGR power reactor. The information is itemized for each reactor component or system and incudes quantity, labor hours, labor cost, material cost, and total costs

  18. Status of international HTGR [high-temperature gas-cooled reactor] development

    International Nuclear Information System (INIS)

    Homan, F.J.; Simon, W.A.

    1988-01-01

    Programs for the development of high-temperature gas-cooled reactor (HTGR) technology over the past 30 years in eight countries are briefly described. These programs have included both government sector and industrial participation. The programs have produced four electricity-producing prototype/demonstration reaactors, two in the United States, and two in the Federal Republic of Germany. Key design parameters for these reactors are compared with the design parameters planned for follow-on commercial-scale HTGRs. The development of HTGR technology has been enhanced by numerous cooperative agreements over the years, involving both government-sponsored national laboratories and industrial participants. Current bilateral cooperative agreements are described. A relatively new component in the HTGR international cooperation is that of multinational industrial alliances focused on supplying commercial-scale HTGR power plants. Current industrial cooperative agreements are briefly discussed

  19. Application of the lines-of-protection concept to the HTGR-SC/C

    International Nuclear Information System (INIS)

    1981-09-01

    The purpose of this document is to present a method for structuring the safety related design and development plans for the HTGR. This method centers on and develops the concept that the HTGR inherently (and by design) provides independent and successive LOPs against potential core related accidents and any resulting public harm. To exemplify the LOP concept and its application to the HTGR, this document identifies some key bases and assumptions, describes the four LOPs selected for the HTGR, identifies the associated safety goals and plant success criteria, and establishes methods for safety research and development prioritization. A task breakdown structure is then described, which in a complete hierarchial fashion can be used to catalog all safety related tasks necessary to demonstrate LOP success as well as catalog safety research areas which cannot be conveniently grouped under the LOPs

  20. Development of plate-fin heat exchanger for intermediate heat exchanger of high-temperature gas cooled reactor. Fabrication process, high-temperature strength and creep-fatigue life prediction of plate-fin structure made of Hastelloy X

    International Nuclear Information System (INIS)

    Mizokami, Yorikata; Igari, Toshihide; Nakashima, Keiichi; Kawashima, Fumiko; Sakakibara, Noriyuki; Kishikawa, Ryouji; Tanihira, Masanori

    2010-01-01

    The helium/helium heat exchanger (i.e., intermediate heat exchanger: IHX) of a high-temperature gas-cooled reactor (HTGR) system with nuclear heat applications is installed between a primary system and a secondary system. IHX is operated at the highest temperature of 950degC and has a high capacity of up to 600 MWt. A plate-fin-type heat exchanger is the most suitable for IHX to improve construction cost. The purpose of this study is to develop an ultrafine plate-fin-type heat exchanger with a finer pitch fin than a conventional technology. In the first step, fabrication conditions of the ultrafine plate fin were optimized by press tests. In the second step, a brazing material was selected from several candidates through brazing tests of rods, and brazing conditions were optimized for plate-fin structures. In the third step, tensile strength, creep rupture, fatigue, and creep-fatigue tests were performed as typical strength tests for plate-fin structures. The obtained data were compared with those of the base metal and plate-fin element fabricated from SUS316. Finally, the accuracy of the creep-fatigue life prediction using both the linear cumulative damage rule and the equivalent homogeneous solid method was confirmed through the evaluation of creep-fatigue test results of plate-fin structures. (author)