WorldWideScience

Sample records for gas-cooled reactor operation

  1. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  2. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  3. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  4. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  5. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

  6. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  7. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Vakilian, M.

    1977-05-01

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  8. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    1974-01-01

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

  9. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    Bastien, D.

    1996-01-01

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  10. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

  11. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    1988-01-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  12. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

    Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

  13. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  14. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  15. Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

    International Nuclear Information System (INIS)

    Halvers, L.J.

    1977-03-01

    An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    Maturana, Roberto Hernan

    2008-01-01

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

  18. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  19. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  20. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  1. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

    2002-01-01

    Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

  2. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

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

  3. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

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

  4. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  5. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

  6. Impact of closed Brayton cycle test results on gas cooled reactor operation and safety

    International Nuclear Information System (INIS)

    Wright, St.A.; Pickard, P.S.

    2007-01-01

    This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas

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

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

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

  8. Overview of environmental control aspects for the gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

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

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  10. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  11. Gas-cooled breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

    1981-01-15

    The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

  12. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  13. Gas-Cooled Reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1978-01-01

    Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

  14. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  15. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  16. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  17. Analysis for RSG-GAS operational characteristics of reactor cooling system

    International Nuclear Information System (INIS)

    Nurhappy, T.

    1998-01-01

    Analysis of operational characteristics of reactor cooling systems (JE01 and PA) is aimed at determining the effects of operation and maintenance patterns to the operational characteristic of the system. Analysis is carried out by virtue of the operating and maintenance data from 1987 to 1997, comprising the operating hours (duration) and data on operating failures of the systems. Results of study show that, either separately or jointly, the operating and maintenance patterns will qualitatively affect the operational characteristic of the systems

  18. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  19. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

  20. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  2. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  3. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  4. Utilization of multi-purpose high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kawada, Osamu; Onuki, Yoshiaki; Wasaoka, Takeshi.

    1974-01-01

    Concerning the utilization of multi-purpose high temperature gas-cooled reactors, the electric power generation with gas turbines is described: features of HTR-He gas turbine power plants; the state of development of He gas turbines; and combined cycle with gas turbines and steam turbines. The features of gas turbines concern heat dissipation into the environment and the mode of load operation. Outstanding work in the development of He gas turbines is that in Hochtemperatur Helium-Turbine Project in West Germany. The power generation with combined gas turbines and steam turbines appears to be superior to that with gas turbines alone. (Mori, K.)

  5. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  6. Periodic reviews of structural integrity of gas-cooled reactors

    International Nuclear Information System (INIS)

    Banks, P.J.; Stokoe, T.Y.; Thomas, D.L.

    1995-01-01

    Nuclear Electric operates 12 gas-cooled reactor power stations which have been in service for between 5 and 30 years. Periodically, comprehensive reviews of the safety cases are carried out for each station. The approach followed in these reviews in respect of structural integrity is outlined with the use of illustrative examples. (author)

  7. The modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  8. The UK gas-cooled reactor programme - Progress report 1988

    International Nuclear Information System (INIS)

    Askew, J.R.

    1989-01-01

    This paper summarises key developments during 1988 on the 26 Magnox reactors and 14 AGRs now operating in the UK. Details are given of long-term safety reviews of the Berkeley and Bradwell Magnox stations which resulted in a decision by CEGB to cease generation at Berkeley but to continue operation at Bradwell. The summary of operating experience with the AGRs concentrates on the completion of construction and successful commissioning of the second generation AGRs at Heysham 2 and Torness. An appended article by John Wilson, Deputy Director of the UKAEA's gas-cooled reactor R and D programme, gives details of the aims and achievements of the programme during 1988. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Scram device for gas-cooled reactor

    International Nuclear Information System (INIS)

    Murakami, Atsushi; Takahashi, Suehiro.

    1989-01-01

    A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

  11. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  12. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail

  13. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  14. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  15. Application of Hastelloy X in gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

  16. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

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

  17. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  18. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  19. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  20. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  1. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  2. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  3. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  4. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  5. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  6. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  7. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  8. Emergency cooling apparatus for reactor

    International Nuclear Information System (INIS)

    Sakaguchi, S.

    1975-01-01

    A nuclear reactor is described which has the core surrounded by coolant and an inert cover gas all sealed within a container, an emergency cooling apparatus employing a detector that will detect cover gas or coolant, particularly liquid sodium, leaking from the container of the reactor, to release a heat exchange material that is inert to the coolant, which heat exchange material is cooled during operation of the reactor. The heat exchange material may be liquid niitrogen or a combination of spheres and liquid nitrogen, for example, and is introduced so as to contact the coolant that has leaked from the container quickly so as to rapidly cool the coolant to prevent or extinguish combustion. (Official Gazette)

  9. Gas-cooled reactor technology: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1981-09-01

    Included are 3358 citations on gas-cooled reactor technology contained in the DOE Energy Data Base for the period January 1978 through June 1981. The citations include reports, journal articles, books, conference papers, patents, and monographs. Corporate, Personal Author, Subject, Contract Number, and Report Number Indexes are provided

  10. Modular high-temperature gas-cooled reactor simulation using parallel processors

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulations routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (operator) interaction. In this paper the benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses

  11. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  12. Frequency and distribution of leakages in steam generators of gas-cooled reactors

    International Nuclear Information System (INIS)

    Bongratz, R.; Breitbach, G.; Wolters, J.

    1988-01-01

    In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm 2 , then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

  13. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1982-01-01

    Towards the second half of a fifty-year time span the market for gas-cooled reactors as sources of high temperature process heat and as highly fuel efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realisation of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel-making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  14. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1983-01-01

    Towards the second half of a 50-year time span the market for gas-cooled reactors as sources of high-temperature process heat and as highly fuel-efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realization of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas, later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  15. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  16. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

  17. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  18. Modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

  1. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  2. Dry cooling tower operating experience in the LOFT reactor

    International Nuclear Information System (INIS)

    Hunter, J.A.

    1980-01-01

    A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

  3. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

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

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

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

  5. Emergency cooling method and system for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1982-01-01

    For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

  6. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  7. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  8. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  9. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

    1981-01-15

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

  10. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.; Cobb, E.C.

    1981-01-01

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

  11. Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1986-01-01

    The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail [fr

  12. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

  13. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

  14. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  15. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-01-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

  16. The Preliminary Study of High Temperature Gas Cooled Reactors (HTGRs) Technology

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Ridzuan Abdul Mutalib

    2015-01-01

    High Temperature Gas Cooled Reactors (HTGRs) have attracted worldwide interest because of their high outlet temperatures, which allow them to be used for applications beyond electricity generation. HTGRs have been built and operated since as far back as the 1970s. Experimental and demonstration reactors of this type have operated in China, Great Britain, Germany, Japan, and the United States of America. This paper is written to share the valuable knowledge and information of HTGRs technology as a mean to enrich peoples understanding of the technology. This paper will present the technological features of HTGRs that allow for a modular design with inherently safe characteristics. (author)

  17. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

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

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

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

  19. Analysis Of The Heat Exchanger Capability At One Line Cooling System Operation Mode Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Kuntoro, Iman

    2000-01-01

    In the frame of minimizing the operation lost of the RSG-GAS reactor, operation using one line cooling system at certain power range is being evaluated. Analysis the performance of cooling system for determining maximum power should be carried out. Analysis was carried out based on heat exchanger calculation using actual operation data. Constraints imposed to the analysis are that inlet cooling system to the reactor core shall be less than 42 o C. The result shows that by using one line of primary and secondary coolant flow of 1780 m exp. 3/hr and 2000 m 3 /hr and secondary coolant temperature from the cooling tower of 38 o C, the primary coolant to the core will be reach 42 o C if reactor operated at power of 16 MW

  20. Refueling system for the gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1980-05-01

    Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

  1. Gas-cooled reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

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

    International Nuclear Information System (INIS)

    1980-12-01

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

  4. Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

    International Nuclear Information System (INIS)

    Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

    2005-01-01

    This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

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

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

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

  6. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

    International Nuclear Information System (INIS)

    Cave, L.; Clarke, A.W.

    1984-01-01

    The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

  8. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  10. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  11. Concept of an inherently-safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-01-01

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  12. Overview of gas cooled reactors' applications with CATHARE

    International Nuclear Information System (INIS)

    Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

    2005-01-01

    Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

  13. Cooling device for reactor suppression pool

    International Nuclear Information System (INIS)

    Togasaki, Susumu; Kato, Kiyoshi.

    1994-01-01

    In a cooling device of a reactor suppression pool, when a temperature of pool water is abnormally increased and a heat absorbing portion is heated by, for example, occurrence of an accident, coolants are sent to the outside of the reactor container to actuates a thermally operating portion by the heat energy of coolants and drive heat exchanging fluids of a secondary cooling system. If the heat exchanging fluids are sent to a cooling portion, the coolants are cooled and returned to the heat absorbing portion of the suppression pool water. If the heat absorbing portion is heat pipes, the coolants are evaporated by heat absorbed from the suppression pool water, steams are sent to the thermally operating portion, then coolants are liquefied and caused to return to the heat absorbing portion. If the thermal operation portion is a gas turbine, the gas turbine is operated by the coolants, and it is converted to a rotational force to drive heat exchanging fluids by pumps. By constituting the cooling portion with a condensator, the coolants are condensed and liquefied and returned to the heat absorbing portion of the suppression pool water. (N.H.)

  14. Optimisation of gas-cooled reactors with the aid of mathematical computers

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H

    1959-04-15

    Reactor optimisation is the task of finding the combination of values of the independent variables in a reactor design producing the lowest cost of electricity. In a gas-cooled reactor the number of independent variables is particularly large and the optimisation process is, therefore, laborious. The present note describes a procedure for performing the entire optimisation procedure with the aid of a mathematical computer in a single operation, thus saving time for the design staff. Detailed equations and numerical constants are proposed for the thermal and cost relations involved. The reactor physics equations, on the other hand are merely stated as general functions of the relevant variables. The task of expressing these functions as detailed equations will be covered by separate documents prepared by the reactor physics department.

  15. Optimisation of gas-cooled reactors with the aid of mathematical computers

    International Nuclear Information System (INIS)

    Margen, P.H.

    1959-04-01

    Reactor optimisation is the task of finding the combination of values of the independent variables in a reactor design producing the lowest cost of electricity. In a gas-cooled reactor the number of independent variables is particularly large and the optimisation process is, therefore, laborious. The present note describes a procedure for performing the entire optimisation procedure with the aid of a mathematical computer in a single operation, thus saving time for the design staff. Detailed equations and numerical constants are proposed for the thermal and cost relations involved. The reactor physics equations, on the other hand are merely stated as general functions of the relevant variables. The task of expressing these functions as detailed equations will be covered by separate documents prepared by the reactor physics department

  16. Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

    International Nuclear Information System (INIS)

    Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

    2003-01-01

    A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

  17. Gas-cooled reactor coolant circulator and blower technology

    International Nuclear Information System (INIS)

    1988-08-01

    In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

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

    1990-02-01

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

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

    International Nuclear Information System (INIS)

    Liu Yu; Dong Yujie

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Jia Qianqian; Qu Ronghong; Zhang Liangju

    2014-01-01

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

  1. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Arai, Kenji.

    1996-01-01

    Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

  2. High temperature gas cooled reactor technology development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-12-01

    The successful introduction of an advanced nuclear power plant programme depends on many key elements. It must be economically competitive with alternative sources of energy, its technical development must assure operational dependability, the support of society requires that it be safe and environmentally acceptable, and it must meet the regulatory standards developed for its use and application. These factors interrelate with each other, and the ability to satisfy the established goals and criteria of all of these requirements is mandatory if a country or a specific industry is to proceed with a new, advanced nuclear power system. It was with the focus on commercializing the high temperature gas cooled reactor (HTGR) that the IAEA's International Working Group on Gas Cooled Reactors recommended this Technical Committee Meeting (TCM) on HTGR Technology Development. Over the past few years, many Member States have instituted a re-examination of their nuclear power policies and programmes. It has become evident that the only realistic way to introduce an advanced nuclear power programme in today's world is through international co-operation between countries. The sharing of expertise and technical facilities for the common development of the HTGR is the goal of the Member States comprising the IAEA's International Working Group on Gas Cooled Reactors. This meeting brought together key representatives and experts on the HTGR from the national organizations and industries of ten countries and the European Commission. The state electric utility of South Africa, Eskom, hosted this TCM in Johannesburg, from 13 to 15 November 1996. This TCM provided the opportunity to review the status of HTGR design and development activities, and especially to identify international co-operation which could be utilized to bring about the commercialization of the HTGR

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  4. The calculating methods of the release of airborne radionuclides to environment during the normal operation of a module high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Yuanzhong

    1993-01-01

    The calculations of the release of radionuclides to environment are the basis of environmental impact assessment during the normal operation of a module high temperature gas-cooled reactor of the Institute of Nuclear Energy Technology, Tsinghua University, China. According to the features of the reactor it is pointed out that only five sources of the airborne radioactive materials released to environment are important. They are: (1) the activation of the air in the reactor cavity; (2) the escape from the primary coolant systems; (3) the release of radioactively contaminated helium from storage tanks; (4) the release of radioactively contaminated helium from the gas evacuation system of fuel load and unload system; (5) the leakage of the vapour from water-steam loop. In accordance with five release sources the calculating methods of radionuclides released to environment are worked out respectively and the respective calculating formulas are derived for the normal operation of the reactor

  5. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    International Nuclear Information System (INIS)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc

  6. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-01

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

  8. Steam generator materials constraints in UK design gas-cooled reactors

    International Nuclear Information System (INIS)

    James, D.W.

    1988-01-01

    A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

  9. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Gotschall, H.L.

    1994-01-01

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

  11. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  14. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    Cleveland, J.; Kupitz, J.

    1992-01-01

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  15. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  16. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

  17. Modeling and Simulation of the Multi-module High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Liu Dan; Sun Jun; Sui Zhe; Xu Xiaolin; Ma Yuanle; Sun Yuliang

    2014-01-01

    The modular high temperature gas-cooled reactor (MHTGR) is characterized with the inherent safety. To enhance its economic benefit, the capital cost of MHTGR can be decreased by combining more reactor modules into one unit and realize the batch constructions in the concept of modularization. In the research and design of the multi-module reactors, one difficulty is to clarify the coupling effects of different modules in operating the reactors due to the shared feed water and main steam systems in the secondary loop. In the advantages of real-time simulation and coupling calculations of different modules and sub-systems, the operation of multi-module reactors can be studied and analyzed to understand the range and extent of the coupling effects. In the current paper; the engineering simulator for the multi-module reactors was realized and able to run in high performance computers, based on the research experience of the HTR-PM engineering simulator. The models were detailed introduced including the primary and secondary loops. The steady state of full power operation was demonstrated to show the good performance of six-module reactors. Typical dynamic processes, such as adjusting feed water flow rates and shutting down one reactor; were also tested to study the coupling effects in multi-module reactors. (author)

  18. Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

    International Nuclear Information System (INIS)

    Cross, M.T.; Wareing, M.I.; Dixon, C.

    1998-01-01

    Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

  19. A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

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

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

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

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

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

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

  2. Nuclear reactor, its cooling facility, nuclear power plant, and method of operating the same

    International Nuclear Information System (INIS)

    Tate, Hitoshi; Tominaga, Kenji; Fujii, Tadashi.

    1993-01-01

    The upper surface of inner structural materials in a container is partitioned by concrete structural walls to form an upper space portion. A pressure relief plate is disposed on the concrete structural walls. If an accident occurs, the pressure relief plate is operated to form a circulation path for a gas to return to the upper space portion again from the upper space portion. The temperature of cooling water in a pressure suppression chamber, that is, a wet well liquid phase portion is elevated by after heat of a reactor core. Evaporated steams transfer from the wet well gas phase portion to the upper space portion passing through pipelines and are mixed with N 2 gas present in the upper space portion. The mixed gas is cooled by a container inner wall cooled by air passing through an air cooling duct, flows downward by way of the pressure relief plate and reaches the wet well gas phase portion again. Since the gases in the upper space circulate by a driving force caused by the after heat, reliability of cooling performance can be improved upon occurrence of an accident without using an active driving force. (I.N.)

  3. LOFC fission product release and circulating activity calculations for gas-cooled reactors

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

    1977-01-01

    The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

  4. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

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

  5. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

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

  6. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  9. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  10. Present state and future prospect of development of high temperature gas-cooled reactors in Japan

    International Nuclear Information System (INIS)

    Sanokawa, Konomo

    1994-01-01

    High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

  11. A three-dimensional thermal and fluid dynamics analysis of a gas cooled subcritical fast reactor driven by a D-T fusion neutron source

    International Nuclear Information System (INIS)

    Angelo, G.; Andrade, D.A.; Angelo, E.; Carluccio, T.; Rossi, P.C.R.; Talamo, A.

    2011-01-01

    Highlights: → A thermal fluid dynamics numerical model was created for a gas cooled subcritical fast reactor. → Standard k-ε model, Eddy Viscosity Transport Equation model underestimates the fuel temperature. → For a conservative assumption, SSG Reynolds stress model was chosen. → Creep strength is the most important parameter in fuel design. - Abstract: The entire nuclear fuel cycle involves partitioning classification and transmutation recycling. The usage of a tokamak as neutron sources to burn spent fuel in a gas cooled subcritical fast reactor (GCSFR) reduces the amount of long-lived radionuclide, thus increasing the repository capacity. This paper presents numerical thermal and fluid dynamics analysis for a gas cooled subcritical fast reactor. The analysis aim to determine the operational flow condition for this reactor, and to compare three distinct turbulence models (Eddy Viscosity Transport Equation, standard k-ε and SSG Reynolds stress) for this application. The model results are presented and discussed. The methodology used in this paper was developed to predict the coolant mass flow rate. It can be applied to any other gas cooled reactor.

  12. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  13. Problems of creating fuel elements for fast gas-cooled reactors working on N2O4-dissociating coolant

    International Nuclear Information System (INIS)

    Nesterenko, V.B.; Zelensky, V.F.; Kolykhan, L.I.; Karpenko, G.V.; Krasnorutsky, V.S.; Isakov, V.P.; Ashikhmin, V.P.; Permyakov, L.N.

    1985-01-01

    A variant of fast gas-cooled reactors is one using dissociating N 2 O 4 nitrogen tetroxide as a coolant. This type of reactors is promising because of great thermal effects of dissociation reactions while heating and recombination while cooling; small latent heat of evaporation; high heat transfer coefficient owing to additional heat transfer in a chemical reaction; high N 2 O 4 density in a gas state at operation parameters. The mentioned advantages give possibility to create a small turbine, heat exchange apparatus and to get high heat production in the active zone. All this opens new ways to increase power plants effectiveness

  14. MHTGR [Modular High-Temperature Gas-Cooled Reactor] technology development plan

    International Nuclear Information System (INIS)

    Homan, F.J.; Neylan, A.J.

    1988-01-01

    This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992

  15. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

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

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Ismail; Sekimoto, Hiroshi

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Chang Oh; Richard Moore; Robert Barner

    2005-01-01

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

  18. Acoustical environment of gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Blevins, R.D.

    1986-01-01

    Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

  19. Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

    International Nuclear Information System (INIS)

    Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

    2007-01-01

    According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

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

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

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

  1. Experience of the remote dismantling of the Windscale advanced gas-cooled reactor and Windscale pile chimneys

    International Nuclear Information System (INIS)

    Wright, E.M.

    1993-01-01

    This paper gives brief descriptions of some of the remote dismantling work and equipment used on two large decommissioning projects: the BNFL Windscale Pile Chimneys Project (remote handling machine, waste packaging machine, remotely controlled excavator, remotely controlled demolition machine) and the AEA Windscale Advanced Gas-cooled Reactor Project (remote dismantling machine, operational waste, bulk removal techniques, semi-remote cutting operations)

  2. Reactors Dynamic analysis Due to Reactivity of The RSG-Gas at One Line Cooling Mode

    International Nuclear Information System (INIS)

    Hastuti, Endiah Puji

    2003-01-01

    In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor power has been determined and steady state and LOFA transient analysis have also been done. To complete those analyses, the reactivity analysis was done by means of a core dynamic and thermal hydraulic code, PARET-ANL. Accident simulation was done. by a ramp reactivity accident due to control rod withdrawal. Reactivity analysis was carried out at two power range i.e. low and high power level, by imposing one line mode reactor protection limits. The results show that technically, the RSG-Gas can be operated safely using one line mode

  3. 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

  4. Study on thermodynamic cycle of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu Xinhe; Yang Xiaoyong; Wang Jie

    2017-01-01

    The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

  5. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  6. The modular high-temperature gas-cooled reactor - a new production reactor

    International Nuclear Information System (INIS)

    Nulton, J.D.

    1990-01-01

    One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

  7. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  8. Safety and licensing of MHTGR [Modular High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

    1987-07-01

    The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error

  9. Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1981-01-15

    The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

  10. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

  11. Evaluation of the RSG-GAS cooling tower performance

    International Nuclear Information System (INIS)

    Suroso

    2003-01-01

    Utilization of RSG-GAS reactor should be operated as efficiently as possible, so that reactor operation planning using one line primary coolant can be anticipated. To analyze the performance of the RSG-GAS cooling tower with one line primary coolant doing by using same data from 10 MW thermal reactor operation. The result were then compare to those achieved using CATHENA code. The results indicated that, for design condition the ratio of water flowrate to air is (L/G) 1.52 and number transfer unit (NTU) is 0.348. For operation condition, the average of L/G and NTU are respectively 1.37 and 0,348. Moreover the results achieved by the code showed that L/G and NTU are respectively 1.35 and 0,302. The performance of cooling tower achieved operation condition and the code results are respectively 91% and 72%. This means that the calculated results are lower than measurement results

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

    International Nuclear Information System (INIS)

    Chang Oh

    2004-01-01

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

  13. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  14. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.) [de

  15. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  16. Simulation of a gas cooled reactor with the system code CATHARE

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Ruby, Alain; Geffraye, Genevieve; Messie, Anne; Saez, Manuel; Tauveron, Nicolas; Widlund, Ola

    2006-01-01

    In recent years the CEA has commissioned a wide range of feasibility studies of future advanced nuclear reactors, in particular gas-cooled reactors (GCR). This paper presents an overview of the use of the thermohydraulics code CATHARE in these activities. Extensively validated and qualified for pressurized water reactors, CATHARE has been adapted to deal also with gas-cooled reactor applications. Rather than branching off a separate GCR version of CATHARE, new features have been integrated as independent options in the standard version of the code, respecting the same stringent procedures for documentation and maintenance. CATHARE has evolved into an efficient tool for GCR applications, with first results in good agreement with existing experimental data and other codes. The paper give an example among the studies already carried out with CATHARE with the case of the Very High Temperature Reactor (VHTR) concepts. Current and future activities for experimental validation of CATHARE for GCR applications are also discussed. Short-term validation activities are also included with the assessment of the German utility Oberhausen II. For the long term, CEA has initiated an ambitious experimental program ranging from small scale loops for physical correlations to component technology and system demonstration loops. (authors)

  17. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  18. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors. Proceedings of a specialists meeting held in Juelich, Germany, 6-8 July 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-15

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs.

  19. Noble gas, binary mixtures for commercial gas-cooled reactor systems

    International Nuclear Information System (INIS)

    El-Genk, M. S.; Tournier, J. M.

    2007-01-01

    Commercial gas cooled reactors employ helium as a coolant and working fluid for the Closed Brayton Cycle (CBC) turbo-machines. Helium has the highest thermal conductivity and lowest dynamic viscosity of all noble gases. This paper compares the relative performance of pure helium to binary mixtures of helium and other noble gases of higher molecular weights. The comparison is for the same molecular flow rate, and same operating temperatures and geometry. Results show that although helium is a good working fluid because of its high heat transfer coefficient and significantly lower pumping requirement, a binary gas mixture of He-Xe with M = 15 gm/mole has a heat transfer coefficient that is ∼7% higher than that of helium and requires only 25% of the number stages of the turbo-machines. The binary mixture, however, requires 3.5 times the pumping requirement with helium. The second best working fluid is He-Kr binary mixture with M = 10 gm/mole. It has 4% higher heat transfer coefficient than He and requires 30% of the number of stages in the turbo-machines, but requires twice the pumping power

  20. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  1. Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

    International Nuclear Information System (INIS)

    Monteith, H.C.

    1978-10-01

    The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

  2. Thermal and flow design of helium-cooled reactors

    International Nuclear Information System (INIS)

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  3. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  4. Heat transfer tests of ribbed surfaces for gas-cooled reactors

    International Nuclear Information System (INIS)

    Klepper, O.H.

    1975-07-01

    The performance of gas-cooled reactors is often limited by the heat transfer in the reactor core. Means for modifying core heat transfer surfaces to improve their performance were investigated. The 0.3-in.-OD stainless steel clad heater rods were photo-etched to produce external ribs 0.006 in. high and 0.12 in. wide with a pitch of 0.072 in. Helical ribs with a helix angle of 37 0 (to promote interchannel flow mixing in a multirod array) were provided on one surface. For comparison purposes, a transversely ribbed surface and a smooth rod were also studied. The test surfaces were 49 in. long with a 24-in. heated region, concentrically arranged inside a smooth 0.602-in.-ID stainless steel tube. Nitrogen gas at pressures up to 400 psig was used as the coolant; the linear heat rating ranged to 6.8 kW/ft at surface temperatures up to 1400 0 F; T/sub w/T/sub b/ varied from 1.2 to 2.4 at Re values up to 450,000. Annulus results were recalculated for rod geometry using two different transformations. Good agreement was observed with applicable literature values. The effectiveness of the surfaces was assessed as the ratio E of the heat transfer coefficients of the roughened rods to that of a smooth rod at the same pumping power. The effectiveness of the spiral ribs ranged from 1.3 to 1.4, and from 1.2 to 1.4 for the transverse ribs, spanning Re values from 60,000 to 400,000. These data include variations introduced by alternate transformation methods that were used to make annulus test results applicable to rod geometry. The surfaces investigated in these tests were considered for fast gas-cooled reactors; however, the range of parameters studied also applies to heat transfer from ribbed rod-type fuel elements in thermal gas-cooled reactors. (U.S.)

  5. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  6. 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

  7. Liquid metal versus gas cooled reactor concepts for a turbo electric powered space vehicle

    International Nuclear Information System (INIS)

    Carre, F.; Proust, E.; Schwartz, J.P.

    1985-01-01

    Recent CNES/CEA prospective studies of an orbit transfer vehicule to be launched by ARIANE V, emphasize the advantage of the Brayton cycle over the thermionics and thermoelectricity, in minimizing the total mass of 100 to 300 kWsub(e) power systems under the constraint specific to ARIANE of a radiator area limited to 95 m 2 . The review of candidate reactor concepts for this application, finally recommends both liquid metal and gas cooled reactors, for their satisfactory adaptation to a reference Brayton cycle and for the available experience from the terrestrial operation of comparable systems

  8. Design and development of gas cooled reactors with closed cycle gas turbines. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-08-01

    Technological advances over the past fifteen years in the design of turbomachinery, recuperators and magnetic bearings provide the potential for a quantum improvement in nuclear power generation economics through the use of the HTGR with a closed cycle gas turbine. Enhanced international co-operation among national gas cooled reactor programmes in these common technology areas could facilitate the development of this nuclear power concept thereby achieving safety, environmental and economic benefits with overall reduced development costs. This TCM and Workshop was convened to provide the opportunity to review and examine the status of design activities and technology development in national HTGR programmes with specific emphasis on the closed cycle gas turbine, and to identify pathways which take advantage of the opportunity for international co-operation in the development of this concept. Refs, figs, tabs

  9. Design and development of gas cooled reactors with closed cycle gas turbines. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    Technological advances over the past fifteen years in the design of turbomachinery, recuperators and magnetic bearings provide the potential for a quantum improvement in nuclear power generation economics through the use of the HTGR with a closed cycle gas turbine. Enhanced international co-operation among national gas cooled reactor programmes in these common technology areas could facilitate the development of this nuclear power concept thereby achieving safety, environmental and economic benefits with overall reduced development costs. This TCM and Workshop was convened to provide the opportunity to review and examine the status of design activities and technology development in national HTGR programmes with specific emphasis on the closed cycle gas turbine, and to identify pathways which take advantage of the opportunity for international co-operation in the development of this concept. Refs, figs, tabs.

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

    International Nuclear Information System (INIS)

    Taketani, K.

    1978-01-01

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

  11. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  12. Development of the design of the High Temperature Gas Cooled Reactor experiment

    International Nuclear Information System (INIS)

    Lockett, G.E.; Huddle, R.A.U.

    1960-01-01

    Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  15. Technology development for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Homan, F.J.; Turner, R.F.

    1989-01-01

    In the USA the Modular High-Temperature Gas-Cooled Reactor is in an advanced stage of design. The related HTGR program areas, the approaches to these programs along with sample results and a description of how these data are used are highlighted in the paper. (author). Figs and tabs

  16. Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

    International Nuclear Information System (INIS)

    Smith, J.; Hart, R.; Lazic, L.

    2009-01-01

    allow steam at the wellhead to be at 8.5 MPa, saturated, despite significant steam pipe lengths. The steam generator system consists of steam generators, pre-heaters and super-heaters, all designed for operation with high temperature helium as a heat transfer medium. This design utilizes worldwide nuclear steam generator as well as fossil-fuel steam generator experience for optimized, reliable performance. The paper describes the safety aspects of the steam generator system, overall layout of the gas-cooled reactor plant and system controls. With this system, the gas-cooled reactor becomes a viable alternative for energy supply in the Oil Sands. (author)

  17. Method of operating a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Lysell, G.

    1975-01-01

    When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

  18. Numerical investigation of heat transfer in high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)

    1995-09-01

    This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

  19. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  20. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

    Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

  1. Emergency reactor cooling device

    International Nuclear Information System (INIS)

    Arakawa, Ken.

    1993-01-01

    An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

  2. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Akiba, Miyuki.

    1996-01-01

    In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

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

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi; Shimokawa, Jun-ichi

    1974-01-01

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

  4. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Zheng Yanhua; Shi Lei; Wang Yan

    2010-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Quadakkers, W.J.; Schuster, H.

    1989-01-01

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

  7. Emergency cooling system for a nuclear reactor in a closed gas turbine plant

    International Nuclear Information System (INIS)

    Frutschi, H.U.

    1974-01-01

    In undisturbed operation of the closed gas turbine plant with compressor stages, reactor, and turbine, a compressor stage driven by a separate motor is following with reduced power. The power input this way is so small that the working medium is just blown through without pressure increase. The compressor stage is connected with the reactor by means of a reactor feedback pipe with an additional cooler and with the other compressor stages by means of a recuperator in the pipe between these and the turbine. In case of emergency cooling, e.g. after the rupture of a pipe with decreasing pressure of the working medium, the feedback pipe is closed short and the additional compressor stage is brought to higher power. It serves as a coolant blower and transfers the necessary amount of working medium to the reactor. The compressor stage is controlled at a constant torque, so that the heat removal from the reactor is adapted to the conditions of the accident. (DG) [de

  8. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  9. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  12. Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

    1997-01-01

    The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

  13. Assessment and status report High-Temperature Gas-Cooled Reactor gas-turbine technology

    International Nuclear Information System (INIS)

    1981-01-01

    Purpose of this report is to present a brief summary assessment of the High Temperature Gas-Cooled Reactor - Gas Turbine (HTGR-GT) technology. The focal point for the study was a potential 2000 MW(t)/800 MW(e) HTGR-GT commercial plant. Principal findings of the study were that: the HTGR-GT is feasible, but with significantly greater development risk than the HTGR-SC (Steam Cycle). At the level of performance corresponding to the reference design, no incremental economic incentive can be identified for the HTGR-GT to offset the increased development costs and risk relative to the HTGR-SC. The relative economics of the HTGR-GT and HTGR-SC are not significantly impacted by dry cooling considerations. While reduced cycel complexity may ultimately result in a reliability advantage for the HTGR-GT, the value of that potential advantage was not quantified

  14. Genetic algorithms and artificial neural networks for loading pattern optimisation of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ziver, A.K. E-mail: a.k.ziver@imperial.ac.uk; Pain, C.C; Carter, J.N.; Oliveira, C.R.E. de; Goddard, A.J.H.; Overton, R.S

    2004-03-01

    A non-generational genetic algorithm (GA) has been developed for fuel management optimisation of Advanced Gas-Cooled Reactors, which are operated by British Energy and produce around 20% of the UK's electricity requirements. An evolutionary search is coded using the genetic operators; namely selection by tournament, two-point crossover, mutation and random assessment of population for multi-cycle loading pattern (LP) optimisation. A detailed description of the chromosomes in the genetic algorithm coded is presented. Artificial Neural Networks (ANNs) have been constructed and trained to accelerate the GA-based search during the optimisation process. The whole package, called GAOPT, is linked to the reactor analysis code PANTHER, which performs fresh fuel loading, burn-up and power shaping calculations for each reactor cycle by imposing station-specific safety and operational constraints. GAOPT has been verified by performing a number of tests, which are applied to the Hinkley Point B and Hartlepool reactors. The test results giving loading pattern (LP) scenarios obtained from single and multi-cycle optimisation calculations applied to realistic reactor states of the Hartlepool and Hinkley Point B reactors are discussed. The results have shown that the GA/ANN algorithms developed can help the fuel engineer to optimise loading patterns in an efficient and more profitable way than currently available for multi-cycle refuelling of AGRs. Research leading to parallel GAs applied to LP optimisation are outlined, which can be adapted to present day LWR fuel management problems.

  15. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  16. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-02-15

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

  17. Development of failure detection system for gas-cooled reactor

    International Nuclear Information System (INIS)

    Feirreira, M.P.

    1990-01-01

    This work presents several kinds of Failure Detection Systems for Fuel Elements, stressing their functional principles and major applications. A comparative study indicates that the method of electrostatic precipitation of the fission gases Kr and Xe is the most efficient for fuel failure detection in gas-cooled reactors. A detailed study of the physical phenomena involved in electrostatic precipitation led to the derivation of an equation for the measured counting rate. The emission of fission products from the fuel and the ion recombination inside the chamber are evaluated. A computer program, developed to simulate the complete operation of the system, relates the counting rate to the concentration of Kr and Xe isotopes. The project of a mock-up is then presented. Finally, the program calculations are compared to experimental data, available from the literature, yielding a close agreement. (author)

  18. High-temperature gas-cooled reactor safety-reliability program plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  19. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  20. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  1. Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

    International Nuclear Information System (INIS)

    Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

    2016-01-01

    Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

  2. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

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

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

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

  4. The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

    International Nuclear Information System (INIS)

    LaBar, Malcolm P.

    2002-01-01

    The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

  5. Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

    2015-10-15

    Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

  6. Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

    International Nuclear Information System (INIS)

    Barlow, S.V.; Wisbey, S.J.; Wood, P.

    1998-01-01

    United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

  7. State of development of gas cooled reactors in the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Mosevitskij, I.S.

    1991-01-01

    In the context of the programme for the development of gas-cooled reactors in the USSR it is reported that pilot plants with VGR-50 MW(el) and VG-400 MW(el) have been developed up to the stage of engineering design and that now the efforts are concentrated on the project of pilot-commercial reactor plant VGM (PCRP VGM) of a modular type with unit thermal power of 200-250 MW. The installation is designed to solve the main scientific and engineering problems of construction of high-temperature gas-cooled reactors, to test equipment components, and to show advantages of the given type of installations having the enhanced safety and capability to generate high-potential heat. The status of work on the PCRP VGM project is described. 3 refs, 1 fig., 1 tab

  8. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  9. Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

    International Nuclear Information System (INIS)

    Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

    2011-01-01

    This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

  10. Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

    2007-09-01

    The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

  11. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van

    2006-01-01

    The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

  12. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  13. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  14. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

    2004-01-01

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  15. Evaluation Of Radioactivity Concentration In The Primary Cooling Water System Of The RSG-GAS During Operation With 30% Silicide Fuels

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Udiyani, P.M.; Setiawanto, Anto

    2001-01-01

    The evaluating radioactivity concentration in the primary cooling water of the RSG-GAS during operation with 30% silicide fuels has been performed. The method of the research is sampling of primary cooling water during operation of the reactor and calculation of its radioactivity concentration. Based on the data obtained from calculation, the identified nuclides in the water are, Mn-56, Sb-124, Sb-122 and Na-24, under the limit of safety value

  16. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  17. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core

  18. The continuous fuel cycle model and the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

    2011-01-01

    The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

  19. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  20. 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

  1. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  2. Gas-cooled reactor commercialization study. Interim report

    International Nuclear Information System (INIS)

    1977-01-01

    This report of the gas-cooled reactor commercialization study completes the technical and cost evaluation portions of this study contract. A final report in December will update the status of the incentive analyses and the issues of commercialization. This study was designed to bring together potential industry participants (utilities and suppliers) to evaluate the commercial potential of the HTGR-SC and to build channels of communication among the participating organizations at the same time that technical, economic and institutional issues were being evaluated. RAMCO, Inc., in suggesting and using this study approach, believes its application extends to any commercialization problem involving multi-party involvement in high capital, intensive, high risk energy technologies

  3. Preliminary Sensitivity Study on Gas-Cooled Reactor for NHDD System Using MARS-GCR

    International Nuclear Information System (INIS)

    Lee, Seung Wook; Jeong, Jae Jun; Lee, Won Jae

    2005-01-01

    A Gas-Cooled Reactor (GCR) is considered as one of the most outstanding tools for a massive hydrogen production without CO 2 emission. Till now, two types of GCR are regarded as a viable nuclear reactor for a hydrogen production: Prismatic Modular Reactor (PMR), Pebble Bed Reactor (PBR). In this paper, a preliminary sensitivity study on two types of GCR is carried out by using MARS-GCR to find out the effect on the peak fuel and reactor pressure vessel (RPV) temperature, with varying the condition of a reactor inlet, outlet temperature, and system pressure for both PMR and PBR

  4. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    1993-02-01

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  5. Adaptation of a robot and tools for dismantling of a gas-cooled reactor

    International Nuclear Information System (INIS)

    Lewis, C.J.A.; Vibert, C.J.T.

    1989-01-01

    This report details the progress on a research programme to develop the techniques and design necessary to facilitate the use of commercially available industrial manipulator systems and cutting tools in nuclear environments, particularly that envisaged whilst decommissioning a gas-cooled reactor. The technology for the type of control and the machines to perform it already exist in the form of industrial-type robots. Development of the techniques for using these machines in a more operator-sensitive environment, together with the requirements of decontamination and radiation tolerance will enable them to be used in place of expensive purpose-built machines at a considerable cost saving. From this work it was possible to highlight the viability and associated costs of modifying a standard manipulator for use in decommissioning operations

  6. Liquid metal reactor cover gas purification and analysis in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Allen, K J [Argonne National Laboratory, EBR-II Division, Idaho Falls, ID (United States); Meadows, G E; Schuck, W J [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

  7. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1987-01-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  8. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

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

    International Nuclear Information System (INIS)

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

    1992-09-01

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

  10. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

    Colak, Ue.; Kadiroghlu, O.K.; Soekmen, C.N.; Schmitt, H.

    1991-01-01

    Large urban areas with unfavourable topographic and meteorological conditions suffer severe air pollution during the winter months. Use of low grade lignites, imported higher quality coal or imported fuel oil are the sources of air pollution in the form of sulphur dioxide, fly ash and soot. Large housing complexes or old and historical locations within the city are in need of pollution free centralized district heating systems. Natural gas imported from the Soviet Union is a solution for this problem. Lack of gas distribution network for high pressure gas within the city is the main bottle-neck for the heating systems utilizing natural gas. Concern of the safety of flammable high pressure gas circulating within the city is another drawback for the natural gas heating systems. Nuclear district heating is an environmentally viable option worth looking into it. Localized urban nuclear heating is an interesting solution for large urban areas with old and historical character. The results of a feasibility study on the HGR application for the Hacettepe University presented here, summarizes the concept of gas-cooled heating reactors specially designed for urban centers. The inherently safe characteristics of the pebble bed heating reactor makes localized urban nuclear heating a viable alternative to other heat sources. An economical analysis of various heat sources with equal power levels is done for the Beytepe campus of Hacettepe University in Ankara. Under special boundary conditions, the price for heat generation can be much lower for nuclear heating with GHR 20 than for hard coal or fuel oil. It is also possible that if the price escalation rate for natural gas exceeds 3%, then nuclear heating with GHR can be more competitive. It is concluded that the nuclear heating of Beytepe campus with a GHR 20 is feasible and economical. (author) 3 figs., 5 refs

  11. Specialists' meeting on heat exchanging components of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components.

  12. Specialists' meeting on heat exchanging components of gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

  13. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  14. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

  15. Requirements for electricity producing gas-cooled reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Schwarz, D.K.J.

    1989-01-01

    The paper describes requirements to a high-temperature gas-cooled reactor from the view-point of a utility in the Federal Republic of Germany. The requirements presented in the paper address different areas including plant size, availability, safety and economics. (author)

  16. Graphite development for gas-cooled reactors in the USA

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

  17. Microscopical examination of carbon deposits formed in the Windscale advanced gas cooled reactor

    International Nuclear Information System (INIS)

    Livesey, D.J.; Chatwin, W.H.; Pearce, J.H.

    1980-12-01

    Methods are described of sampling and examining carbon deposits on fuel cladding in the Windscale advanced gas-cooled reactor. Deposition is observed on fuel cladding in both the reactor core and experimental loops in carbon dioxide coolants containing various amounts of carbon monoxide and methane. Deposit distribution over the cladding surface indicated that nucleation is dependent on local surface conditions. Microscopical examination showed that deposit thickness increases by carbon filament growth into the coolant gas stream and that the process can be markedly influenced by metallic impurities. There is evidence that nickel can play a particularly significant role in deposition in loop experiments but similar effects have not been observed in the reactor core. (author)

  18. Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

    International Nuclear Information System (INIS)

    Malo, J.Y.; Alpy, N.; Bentivoglio, F.

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

  19. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

  20. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    1985-01-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  1. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  2. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

    1996-01-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  3. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  4. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  5. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

    International Nuclear Information System (INIS)

    Boudouresque, B.; Courcon, P.; Lestiboubois, G.

    1964-01-01

    The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr

  6. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies

  7. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  8. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  9. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  10. Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

    2014-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

  11. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1986-09-01

    Two sodium cooled reactors are currently being operated in the United States of America for the US Department of Energy. These are Experimental Breeder Reactor 11, EBR-11, and the Fast Flux Test Facility, FFTF. EBR-11 is located near Idaho Falls, Idaho, and the FFTF is near Richland, Washington. These reactors are currently engaged in a wide range of testing including fuels and materials tests, and plant system performance and safety development. The US DOE program also includes designs of a next generation sodium cooled power reactor. The FFTF and EBR-11 communities are providing input to these designs. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  12. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  13. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-09-01

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  14. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  15. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  16. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  17. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  18. Eighth meeting of the International Working Group on Gas-Cooled Reactors Vienna, 30 January - 1 February 1989. Summary report. Part 2

    International Nuclear Information System (INIS)

    1989-12-01

    The Eighth Meeting of the IAEA International Working Group on Gas-Cooled Reactors was held in Vienna, Austria, from 30 January - 1 February, 1989. The Summary Report (Part II) contains the papers which review the national programmes in the field of Gas-Cooled Reactors and other presentations at the Meeting. Refs, figs and tabs

  19. AREVA Modular Steam Cycle – High Temperature Gas-Cooled Reactor Development Progress

    International Nuclear Information System (INIS)

    Lommers, L.; Shahrokhi, F.; Southworth, F.; Mayer, J. III

    2014-01-01

    The AREVA Steam Cycle – High Temperature Gas-Cooled Reactor (SCHTGR) is a modular graphite-moderated gas-cooled reactor currently being developed to support a wide variety of applications including industrial process heat, high efficiency electricity generation, and cogeneration. It produces high temperature superheated steam which makes it a good match for many markets currently dependent on fossil fuels for process heat. Moreover, the intrinsic safety characteristics of the SC-HTGR make it uniquely qualified for collocation with large industrial process heat users which is necessary for serving these markets. The NGNP Industry Alliance has selected the AREVA SC-HTGR as the basis for future development work to support commercial HTGR deployment. This paper provides a concise description of the SC-HTGR concept, followed by a summary of recent development activities. Since this concept was introduced, ongoing design activities have focused primarily on confirming key system capabilities and the suitability for potential future markets. These evaluations continue to confirm the suitability of the SC-HTGR for a variety of potential applications that are currently dependent on fossil fuels. (author)

  20. Parametric study of sodium aerosols in the cover-gas space of sodium-cooled reactors

    International Nuclear Information System (INIS)

    Sheth, A.

    1975-03-01

    A mathematical model has been developed to describe the behavior of sodium aerosols in the cover-gas space of a sodium-cooled reactor. A review of the literature was first made to examine methods of aerosol generation, mathematical expressions representing aerosol behavior, and pertinent experimental investigations of sodium aerosols. In the development of the model, some terms were derived from basic principles and other terms were estimated from available correlations. The model was simulated on a computer, and important parameters were studied to determine their effects on the overall behavior of sodium aerosols. The parameters studied were sodium pool temperature, source and initial size of particles, film thickness at the sodium pool/cover gas interface, wall plating parameters, cover-gas flow rate, and type of cover gas (argon and helium). The model satisfactorily describes the behavior of sodium aerosol in argon, but not in helium. Possible reasons are given for the failure of the model with helium, and further experimental work is recommended. The mathematical model, with appropriate modifications to describe the behavior of sodium aerosols in helium, would be very useful in designing traps to remove aerosols from the cover gas of sodium-cooled reactors. (U.S.)

  1. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  2. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  3. Gas-cooled reactor commercialization study: introduction scenario and commercialization analyses for process heat applications. Final report, July 8, 1977--November 30, 1977

    International Nuclear Information System (INIS)

    1977-12-01

    This report identifies and presents an introduction scenario which can lead to the operation of High Temperature Gas Cooled Reactor demonstration plants for combined process heat and electric power generation applications, and presents a commercialization analysis relevant to the organizational and management plans which could implement a development program

  4. Gas-cooled reactor commercialization study: introduction scenario and commercialization analyses for process heat applications. Final report, July 8, 1977--November 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    This report identifies and presents an introduction scenario which can lead to the operation of High Temperature Gas Cooled Reactor demonstration plants for combined process heat and electric power generation applications, and presents a commercialization analysis relevant to the organizational and management plans which could implement a development program.

  5. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  6. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  8. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  9. Design and development of gas turbine high temperature reactor 300 (GTHTR300)

    International Nuclear Information System (INIS)

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

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) started design and development of the high temperature gas cooled reactor with a gas turbine electric generation system, GTHTR300, in April 2001. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japan's first high temperature gas cooled reactor (HTTR: High Temperature Engineering Test Reactor) so that many R and Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost. This paper describes the unique design features focusing on the safety design, reactor core design and gas turbine system design together with a preliminary result of the safety evaluation carried out for a typical severe event. This study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  10. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

    Penfield, S.R. Jr.; Rodwell, E.

    2001-01-01

    The need to avoid contamination of the primary system, along with other perceived advantages, has led to the selection of electromagnetic bearings (EMBs) in most ongoing commercial-scale gas cooled reactor (GCR) designs. However, one implication of magnetic bearings is the requirement to provide backup support to mitigate the effects of failures or overload conditions. The demands on these auxiliary or 'catcher' bearings have been substantially escalated by the recent development of direct Brayton cycle GCR concepts. Conversely, there has been only limited directed research in the area of auxiliary bearings, particularly for vertically oriented turbomachines. This paper explores the current state-of-the-art for auxiliary bearings and the implications for current GCR designs. (author)

  11. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Chen Yihua; Wang Jie; Zhang Zuoyi

    2003-01-01

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

  14. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  15. Investigation of thermohydraulic parameter variations in operating conditions of Bilibino NPP CPS cooling circuit

    International Nuclear Information System (INIS)

    Baranaev, Yu.D.; Koz'menkova, V.V.; Parafilo, L.M.

    2015-01-01

    In consequence of activities on uncovering the reasons for through faults formation in cooling tubes of reactor control and protection system (CPS) channels of Bilibino-2 reactor the conclusion is made that corrosion failure development takes place against the backdrop of periodic increase of total moisture in reactor space at transient and standby modes at top of cooling tubes of CPS channels. Formation of corrosion defects in cooling tubes of four CPS channels of unit 2 in 2011-2012 is specific factor for this plant unit and do not effect on operation of other plant units. It is pointed out that ingress of moisture into gas system of the reactor is the critical factor providing integrity of structure elements of EhPG-6 reactor core cooling system. This fact agrees nicely with the results obtained during operation AM reactor of the First NPP [ru

  16. A UKAEA review of gas-cooled reactors in the United Kingdom

    International Nuclear Information System (INIS)

    Heath, E.C.; Knowles, A.N.

    1983-01-01

    The commercial use of nuclear power for electrical generation commenced in the UK in the 1950s with the Calder Hall reactors. Based on this concept, eighteen commercial reactor units, with two further units outside the UK, were constructed and have been in operation for periods ranging from 10 to 19 years. The paper reviews this experience mainly from the aspects of safety and the achieved costs, which compare favourably with current figures for fossil fired generation. The further development of the gas-cooled system in the UK commenced with the construction of the Windscale AGR, which came into operation in 1962. This led to the ordering of 14 large commercial AGR units, 4 of which have been in service since 1976, 6 are at an advanced stage of construction and 4 are at an early stage of construction. The paper reviews the main safety features of the AGR and considers the costs, taking achieved costs for the units which are in service and a combination of historical costs and projected costs for the units under construction. Again a clear advantage over fossil fuelled stations is shown. The paper also includes a preliminary account of the use of the prototype AGR at Windscale for the series of experiments concerning plateout, over-temperature in the fuel and simulated fault transients in the core which were carried out earlier in 1981. (author)

  17. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  1. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  2. Natural convection in closed vertical cylinders with particular reference to gas cooled reactor standpipes

    International Nuclear Information System (INIS)

    Spence, I.D.

    1975-09-01

    The access to the core for fuel assemblies and control rods of the Advanced Gas Cooled Reactor is through the top cap by means of standpipes. The standpipe is essentially a cylindrical, vertical tube with cooled side wall, closed upper end and an orifice at the lower end which is exposed to the hot core fluid. This creates confined natural convection flow in the empty standpipe and this is the subject of this thesis. The investigation is carried out using analytical and experimental methods. For the analytical work, solution of laminar and turbulent flow is attempted using finite-difference computer techniques. The laminar flow performance is evaluated using two different finite-difference procedures, and the results are compared to each other and to existing analytical and experimental results for the open thermosyphon with cool inflow and hot sidewall, i.e. the complementary problem to the present one. For turbulent flow a two equation turbulence model is employed which provides transport equations for the kinetic energy of turbulence and its dissipation rate. The experimental rig is a full scale replica of the Advanced Gas Cooled Reactor control rod mechanism standpipe. Carbon dioxide and helium are used as the working fluids for the series of tests. (author)

  3. Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Billone, M.C.; Rest, J.

    1982-03-01

    The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects

  4. R + D work on gas-cooled breeder development

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

    1978-01-01

    The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

  5. High-temperature gas-cooled reactor steam cycle/cogeneration application study update

    International Nuclear Information System (INIS)

    1981-09-01

    Since publication of a report on the application of a High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) plant in December of 1980, progress has continued on application related activities. In particular, a reference plant and an application identification effort has been performed, a variable cogeneration cycle balance-of-plant design was developed and an updated economic analysis was prepared. A reference HTGR-SC/C plant size of 2240 MW(t) was selected, primarily on the basis of 2240 MW(t) being in the mid-range of anticipated application needs and the availability of the design data from the 2240 MW(t) Steam Cycle/Electric generation plant design. A variable cogeneration cycle plant design was developed having the capability of operating at a range of process steam loads between the reference design load (full cogeneration) and the no process steam load condition

  6. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  7. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  8. Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

    International Nuclear Information System (INIS)

    Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

    1976-11-01

    During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C 2 Cl 4 ) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed

  9. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

    2014-01-01

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  11. Single phase and two phase erosion corrosion in broilers of gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrison, G.S.; Fountain, M.J.

    1988-01-01

    Erosion-corrosion is a phenomenon causing metal wastage in a variety of locations in water and water-steam circuits throughout the power generation industry. Erosion-corrosion can occur in a number of regions of the once-through boiler designs used in the later Magnox and AGR type of gas cooled nuclear reactor. This paper will consider two cases of erosion-corrosion damage (single and two phase) in once through boilers of gas cooled reactors and will describe the solutions that have been developed. The single phase problem is associated with erosion-corrosion damage of mild steel downstream of a boiler inlet flow control orifice. With metal loss rates of up to 1 mm/year at 150 deg. C and pH in the range 9.0-9.4 it was found that 5 μg/kg oxygen was sufficient to reduce erosion-corrosion rates to less than 0.02 mm/year. A combined oxygen-ammonia-hydrazine feedwater regime was developed and validated to eliminate oxygen carryover and hence give protection from stress corrosion in the austenitic section of the AGR once through boiler whilst still providing erosion-corrosion control. Two phase erosion-corrosion tube failures have occurred in the evaporator of the mild steel once through boilers of the later Magnox reactors operating at pressures in the range 35-40 bar. Rig studies have shown that amines dosed in the feedwater can provide a significant reduction in metal loss rates and a tube lifetime assessment technique has been developed to predict potential tube failure profiles in a fully operational boiler. The solutions identified for both problems have been successfully implemented and the experience obtained following implementation including any problems or other benefits arising from the introduction of the new regimes will be presented. Methods for monitoring and evaluating the efficiency of the solutions have been developed and the results from these exercises will also be discussed. Consideration will also be given to the similarities in the metal loss

  12. Single phase and two phase erosion corrosion in broilers of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, G S; Fountain, M J [Operational Engineering Division (Northern Area), Central Electricity Generating Board, Manchester (United Kingdom)

    1988-07-01

    Erosion-corrosion is a phenomenon causing metal wastage in a variety of locations in water and water-steam circuits throughout the power generation industry. Erosion-corrosion can occur in a number of regions of the once-through boiler designs used in the later Magnox and AGR type of gas cooled nuclear reactor. This paper will consider two cases of erosion-corrosion damage (single and two phase) in once through boilers of gas cooled reactors and will describe the solutions that have been developed. The single phase problem is associated with erosion-corrosion damage of mild steel downstream of a boiler inlet flow control orifice. With metal loss rates of up to 1 mm/year at 150 deg. C and pH in the range 9.0-9.4 it was found that 5 {mu}g/kg oxygen was sufficient to reduce erosion-corrosion rates to less than 0.02 mm/year. A combined oxygen-ammonia-hydrazine feedwater regime was developed and validated to eliminate oxygen carryover and hence give protection from stress corrosion in the austenitic section of the AGR once through boiler whilst still providing erosion-corrosion control. Two phase erosion-corrosion tube failures have occurred in the evaporator of the mild steel once through boilers of the later Magnox reactors operating at pressures in the range 35-40 bar. Rig studies have shown that amines dosed in the feedwater can provide a significant reduction in metal loss rates and a tube lifetime assessment technique has been developed to predict potential tube failure profiles in a fully operational boiler. The solutions identified for both problems have been successfully implemented and the experience obtained following implementation including any problems or other benefits arising from the introduction of the new regimes will be presented. Methods for monitoring and evaluating the efficiency of the solutions have been developed and the results from these exercises will also be discussed. Consideration will also be given to the similarities in the metal loss

  13. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  14. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  15. Description of the magnox type of gas cooled reactor (MAGNOX)

    International Nuclear Information System (INIS)

    Jensen, S.E.; Nonboel, E.

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO 2 ) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  16. Modeling and Application of Pneumatic Conveying for Spherical Fuel Element in Pebble-Bed Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhou Shuyong; Wang Junsan; Wang Yuding; Cai Ruizhong; Zhang Xuan; Cao Jianting

    2014-01-01

    The fuel handling system is an important system for on-load refueling in pebble-bed modular high-temperature gas-cooled reactor. A dynamic model of pneumatic conveying for spherical fuel element in fuel handling system was established to describe the pneumatically conveying process. The motion characteristics of fuel elements in pipeline and the effect of fuel elements on gas velocity were studied using the model. The results show that the theoretical analyses are consistent with the experimental. The research has been used in developing full scope simulator for pebble-bed modular high-temperature gas-cooled reactor, also provides references for the design and optimization of the fuel handling system. (author)

  17. 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

  18. Current design efforts for the gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K.D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  19. Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

    1976-11-01

    During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C/sub 2/Cl/sub 4/) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed.

  20. Digital Information Platform Design of Fuel Element Engineering For High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Du Yuwei

    2014-01-01

    This product line provide fuel element for high temperature gas-cooled reactor nuclear power plant which is being constructed in Shidao bay in Shandong province. Its annual productive capacity is thirty ten thousands fuel elements whose shape is spherical . Compared with pressurized water fuel , this line has the feature of high radiation .In order to reduce harm to operators, the comprehensive information platform is designed , which can realize integration of automation and management for plant. This platform include two nets, automation net using field bus technique and information net using Ethernet technique ,which realize collection ,control, storage and publish of information.By means of construction, automatization and informatization of product line can reach high level. (author)

  1. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  2. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  3. The development of advanced gas cooled reactor iodine adsorber systems

    International Nuclear Information System (INIS)

    Meddings, P.

    1986-01-01

    Advanced Gas Cooled Reactors (AGRs) are provided with plants to process the carbon dioxide coolant prior to its discharge to atmosphere. Included in these are beds of granular activated charcoal, contained within a suitable pressure vessel, through which the high pressure carbon dioxide is passed for the purpose of retaining iodine and iodine-containing compounds. Carry-over carbon dust from the adsorption beds was identified during active in-situ commissioning testing, radio-iodine being transported with the particulate material due to gross disturbance of the adsorber carbon bed and displacement of the vessel internals. The methods used to identify the causes of the problems and find solutions are described. A development programme for the Heysham-2 and Torness reactors iodine adsorber units was set up to identify a method of de-dusting granular charcoal and develop it for full-scale use, of assess the effect under conditions of high gas density of approach velocity on charcoal fines production and to establish the pressure drop characteristics of a packed granular bed and to develop an effective design of inlet gas diffuser manifold to ensure an acceptable velocity distribution. This has involved the construction of a small scale high pressure carbon dioxide rig and development of an air flow model. This work is described. (UK)

  4. The gas-cooled high temperature reactor: perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

  5. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  6. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  7. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    1984-12-01

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

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

  9. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  10. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  11. Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

    2016-09-15

    The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

  12. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    International Nuclear Information System (INIS)

    Ball, S.J.

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

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

    Science.gov (United States)

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

    2011-03-01

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

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

    International Nuclear Information System (INIS)

    Zhu Yan; Xu Yuanhui

    1998-01-01

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

  15. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure

  16. Status of national gas cooled reactor programmes

    International Nuclear Information System (INIS)

    1991-08-01

    This report has been compiled as a central source of summary-level information on the present status of High Temperature Gas-Cooled Reactor (HTGR) programmes in the world and on future plans for the continued development and deployment of HTGRs. Most of the information concerns the programmes in the United States, Germany, Japan and the Soviet Union, countries that have had large programmes related to HTGR technology for several years. Summary-level information is also provided in the report on HTGR-related activities in several other countries who either have an increasing interest in the technology and/or who are performing some development efforts related to HTGR technology. The report contains a summary-level update on the MAGNOX and AGR programmes. This is the twelfth issue of the document, the first of which was issued in March, 1979. The report has been prepared in the IAEA Nuclear Power Technology Development Section. Figs and tabs

  17. Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

    International Nuclear Information System (INIS)

    Parks, G.T.

    1987-01-01

    A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

  18. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  19. Sodium-cooled reactors, objectives, achieved technical state and development trends

    International Nuclear Information System (INIS)

    Wolff, U.

    1988-01-01

    The use of fossil fuels to cover the future world-wide energy demand alone would rapidly deplete these ressources, especially oil and gas. Today's knowledge suggests the enhanced exploitation of solar energy, nuclear fusion and the application of uranium in sodium-cooled breeder reactors as the alternative energies offering a great potential. The sodium-cooled reactor outdistances the other options in terms of development. Its technical feasibility and safe operation have been verified and its profitability appears to be possible when using today's technology. The verification of its profitability while maintaining a high safety level is the overriding task for the future. The paper discusses corresponding activities in the USA, the USSR, Japan and Western Europe. (orig.) [de

  20. Operation method and operation control device for emergency core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

    1996-05-07

    The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

  1. Contributions to the neutronic analysis of a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Reyes-Ramirez, Ricardo; Francois, Juan-Luis; Reinking-Cejudo, Arturo G.

    2011-01-01

    Highlights: → Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. → Fuel lattice and core criticality calculations were done. → A higher Doppler coefficient than coolant density coefficient. → Zirconium carbide is a better reflector than silicon carbide. → Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the

  2. The variation of particle gas-borne concentration with time in a gas cooled reactor

    International Nuclear Information System (INIS)

    Reed, J.; Hall, D.; Reeks, M.W.

    1985-01-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  3. The variation of particle gas-borne concentration with time in a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reed, J; Hall, D; Reeks, M W [Central Electricity Generating Board, Berkeley Nuclear Laboratories (United Kingdom)

    1985-07-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  4. 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.)

  5. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

  6. Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

    International Nuclear Information System (INIS)

    Rimpault, G.; Sugino, K.; Hayashi, H.

    2005-01-01

    In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

  7. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  8. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

    International Nuclear Information System (INIS)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

  9. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-15

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  10. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    2013-04-01

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  11. Specialists' meeting on gas-cooled reactor core and high temperature instrumentation, Windermere, UK, 15-17 June 1982. Summary report

    International Nuclear Information System (INIS)

    1982-09-01

    The Specialists' Meeting on ''Gas-Cooled Reactor Core and High Temperature Instrumentation'' was held at the Beech Hill Hotel, Windermere in England on June 15-17 1982. The meeting was sponsored by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors and was hosted by the Windscale Nuclear Power Development Laboratories of the UKAEA. The meeting was attended by 43 participants from Belgium, France, Federal Republic of Germany, Japan, United Kingdom of Great Britain and Northern Ireland and the United States of America. The objective of the meeting was to provide a forum, both formal and informal, for the exchange and discussion of technical information relating to instrumentation being used or under development for the measurement of core parameters, neutron flux, temperature, coolant flow etc. in gas cooled reactors. The technical part of the meeting was divided into five subject sessions: (A) Temperature Measurement (B) Neutron Detection Instrumentation (C) HTR Instrumentation - General (D) Gas Analysis and Failed Fuel Detection (E) Coolant Mass Flow and Leak Detection. A total of twenty-five papers were presented by the participants on behalf of their organizations during the meeting. A programme of the meeting and list of participants are given in appendices to this report

  12. 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)

  13. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  14. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  15. Variable electricity and steam from salt, helium and sodium cooled base-load reactors with gas turbines and heat storage - 15115

    International Nuclear Information System (INIS)

    Forsberg, C.; McDaniel, P.; Zohuri, B.

    2015-01-01

    Advances in utility natural-gas-fired air-Brayton combed cycle technology is creating the option of coupling salt-, helium-, and sodium-cooled nuclear reactors to Nuclear air-Brayton Combined Cycle (NACC) power systems. NACC may enable a zero-carbon electricity grid and improve nuclear power economics by enabling variable electricity output with base-load nuclear reactor operations. Variable electricity output enables selling more electricity at times of high prices that increases plant revenue. Peak power is achieved using stored heat or auxiliary fuel (natural gas, bio-fuels, hydrogen). A typical NACC cycle includes air compression, heating compressed air using nuclear heat and a heat exchanger, sending air through a turbine to produce electricity, reheating compressed air, sending air through a second turbine, and exhausting to a heat recovery steam generator (HRSG). In the HRSG, warm air produces steam that is used to produce added electricity. For peak power production, auxiliary heat (natural gas, stored heat) is added before the air enters the second turbine to raise air temperatures and power output. Like all combined cycle plants, water cooling requirements are dramatically reduced relative to other power cycles because much of the heat rejection is in the form of hot air. (authors)

  16. Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO{sub 2} gas

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yeon-Gun [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear and Energy Engineering, Jeju National University, 102 Jejudaehakno, Jeju-si 690-756, Jeju (Korea, Republic of); Park, Il-Woong [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Lee, Dong Won [Nuclear Fusion Engineering Development Center, Korea Atomic Energy Research Institute, Daedeokdaero 989 beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Eung-Soo, E-mail: kes7741@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2015-11-15

    Helium (He) cooling requires large circulation power to remove high heat from plasma side and nuclear heating by high energy neutron in fusion reactors due to its low density. Based on the recent findings that the heat transfer capability of the light gas can be enhanced by mixing another heavier gas, this study adds CO{sub 2} to a reference helium coolant and evaluates the cooling performance of the binary mixture for various compositions. To assess the cooling performance, computational fluid dynamic (CFD) analyses on the KO HCML (Korea Helium Cooled Molten Lithium) TBM are conducted. As a result, it is revealed that the binary mixing of helium, which has favorable thermophysical properties but the density, with a heavier noble gas or an unreactive gas significantly reduces the required circulation power by an order of magnitude with meeting the thermal design requirements. This is attributed to the fact that the density can be highly increased with small amount of a heavier gas while other gas properties are kept relatively comparable. The optimal CO{sub 2} mole fraction is estimated to be 0.4 and the circulation power, in this case, can be reduced to 13% of that of pure helium. This implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of the proposed binary mixing.

  17. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  18. Modelling of plate-out under gas-cooled reactor (GCR) accident conditions

    International Nuclear Information System (INIS)

    Taig, A.R.

    1981-01-01

    The importance of plate-out in mitigating consequences of gas-cooled reactor accidents, and its place in assessing these consequences, are discussed. The data requirements of a plate-out modelling program are discussed, and a brief description is given of parallel work programs on thermal/hydraulic reactor behaviour and fuel modelling, both of which will provide inputs to the plate-out program under development. The representation of a GCR system used in SRD studies is presented, and the equations governing iodine adsorption, desorption and transport round the circuit are derived. The status of SRD's plate-out program is described, and the type of sensitivity studies to be undertaken with the partially-developed computer program in order to identify the most useful lines for future research is discussed. (author)

  19. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  20. Thermodynamic data for selected gas impurities in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.C.

    1976-12-01

    The literature of thermodynamic data for selected fission-product species is reviewed and supplemented in support of complex chemical equilibrium calculations applied to fission-product distributions in the primary coolant of high-temperature gas-cooled reactors. Thermodynamic functions and heats and free energies of formation are calculated and tabulated to 3000 0 K for CsI (s,l,g), Cs 2 I 2 (g), CH 3 I(g), COI 2 (g), and CsH(g). 79 references

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

    International Nuclear Information System (INIS)

    Marmier, Alain

    2012-01-01

    This thesis covers 3 fundamental aspects of High Temperature Reactor (HTR) performance: fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R and D. The HTR concept features important inherent and passive safety characteristics: high thermal inertia and good thermal conductivity of the core; a negative Doppler coefficient; high quality of fuel elements and low power density. These features keep the core temperature within safe boundaries and minimise fission product release, even in case of severe accidents. The Very High Temperature reactor (VHTR) is based on the same safety concept as the initial HTR, but it aims at offering better economy with a higher reactor outlet temperature (and thus efficiency) and a high fuel discharge burn-up (and thus better sustainability). The inherent safety features of HTR have been demonstrated in small pebble-bed reactors in practice, but have to be replicated for reactors with industrially relevant size and power. An increase of the power density (in order to increase the helium coolant outlet temperature) leads to higher fuel temperatures and therefore higher fuel failure probability. The core of a pebble-bed reactor consists of 6 cm diameter spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. These pebbles contain thousands of 1 mm diameter fuel particles baked into a graphite matrix. These fuel particles, in turn, consist of a fuel kernel with successive coatings of pyrocarbon and silicon carbide layers. The coating layers are designed to contain the fission products that build up during operation of the reactor. The feasibility and performance of the fuel requires experimental verification in view of fuel qualification and licensing. For HTR fuel, the required test string comprises amongst others

  2. Application of noise analysis on the operation of the RSG-GAS reactor

    International Nuclear Information System (INIS)

    Surian P, Tukiran S

    1999-01-01

    The RSG-GAS reactor has been operating for radioisotope production and experiments so that it is necessary to perform analysis of the reactor operation. analysis was done based on reactor noise experiment. Neutron noise at low and high power of the RSG-GAS has been analyzed using time and frequency domain with aim to determine the safety of reactor operation. The safety of reactor operation based on two parameters as, prompt neutron decay constant and decay ratio. The parameters are useful for reactor operation, so it is necessary to determine accurately. For determining prompt neutron decay constant, neutron density in the core of reactor which operated at 10 k W, was collected by using Fission Chamber detectors (FC). Based on power spectral density (PSD) was achieved break frequency about 23 Hz, so that the prompt neutron decay constant is about 151 sec -1 . While at at high power 20 MW, neutron density was collected by using Compensated Ionization Chamber (CIC) detector. The result at high power showed that there is reactivity effect in the core because of fluctuation in temperature and density of the coolant, and the decay ratio of 0.20, showed that the reactor is still operation in stable

  3. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  4. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

  5. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

    Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

    1997-01-01

    The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

  6. Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2005-01-01

    The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine

  7. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  8. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    International Nuclear Information System (INIS)

    Chang Oh

    2006-01-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

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

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2004-01-01

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

  10. Proposal of a fast gas-cooled reactor using transuranics

    International Nuclear Information System (INIS)

    Macedo, Anderson Altair Pinheiro de

    2016-01-01

    In the last two decades, nations that have invested in research and energy generation through nuclear source have devoted part of their efforts in developing new technologies for nuclear reactors. Part of this investment focuses on new material testing, particularly regarding new fuels. In a world view that breaths sustainability, the reprocess and reuse of spent fuel from conventional reactors comes alive in nuclear technology, presenting itself as a real alternative of energy source for the latest generation of reactors. Different concepts of fourth generation reactors have been proposed and must meet some basic requirements, such as: extended burnup, improvement of passive safety, better radioactive waste management, possibility to use reprocessed fuel and proliferation resistance. In this context, the GFR (Gas-cooled Fast Reactor) is one of the future promises, presenting satisfactory neutronic results on the use of type of fuel (U, Pu) C. In the present work, the fuel of a traditional GFR reactor that uses (U, Pu)C was sub was replaced by a transuranic reprocessed fuel (TRU), obtained by non-proliferation reprocessing technology. The UO 2 fuel initially enriched by 3.1% was burned in a standard PWR, with full burn of 33,000 MWd/T. Afterward it was left in a pool for 5 years and finally reprocessed by UREX + method. Two fuels were studied and evaluated, one diluted with depleted uranium (U, TRU)C, and the other diluted in thorium (Th, TRU)C. Assessments were done in steady state and as well as during burning and were compared with results obtained using the standard fuel, (U, Pu) C. The outcome shows that the use of TRU as a fuel, in GFR type reactors, is a real possibility. The research was done using the SCALE 6.0 code modules. (author)

  11. Development of gas-cooled fast reactor and its thermo-hydraulics

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi

    1977-10-01

    Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  13. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  14. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  15. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  16. A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu, Xinhe; Yang, Xiaoyong; Wang, Jie

    2017-01-01

    Highlights: • The features of three different power generation schemes, including closed Brayton cycle, non-reheating combined cycle and reheating combined cycle, coupled with high temperature gas-cooled reactor (HTGR) were investigated and compared. • The effects and mechanism of reactor core outlet temperature, compression ratio and other key parameters over cycle characteristics were analyzed by the thermodynamic models.. • It is found that reheated combined cycle has the highest efficiency. Reactor outlet temperature and main steam parameters are key factors to improve the cycle’s performance. - Abstract: With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900 °C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900 °C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.

  17. Characterization of effluents from a high-temperature gas-cooled reactor fuel refabrication plant

    International Nuclear Information System (INIS)

    Judd, M.S.; Bradley, R.A.; Olsen, A.R.

    1975-12-01

    The types and quantities of chemical and radioactive effluents that would be released from a reference fuel refabrication facility for the High-Temperature Gas-Cooled Reactor (HTGR) have been determined. This information will be used to predict the impact of such a facility on the environment, to identify areas where additional development work needs to be done to further identify and quantify effluent streams, and to limit effluent release to the environment

  18. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

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

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

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

  20. Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

    International Nuclear Information System (INIS)

    Kotake, Syoji; Nishikawa, Akira

    2005-01-01

    Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

  1. A review of reactor physics uncertainties and validation requirements for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Baxter, A.M.; Lane, R.K.; Hettergott, E.; Lefler, W.

    1991-01-01

    The important, safety-related, physics parameters for the low-enriched Modular High-Temperature gas-Cooled Reactor (MHTGR) such as control rod worth, shutdown margins, temperature coefficients, and reactivity worths, are considered, and estimates are presented of the uncertainties in the calculated values of these parameters. The basis for the uncertainty estimate in several of the important calculated parameters is reviewed, including the available experimental data used in obtaining these estimates. Based on this review, the additional experimental data needed to complete the validation of the methods used to calculate these parameters is presented. The role of benchmark calculations in validating MHTGR reactor physics data is also considered. (author). 10 refs, 5 figs, 3 tabs

  2. Benchmark Analysis Of The High Temperature Gas Cooled Reactors Using Monte Carlo Technique

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huda, M.Q.

    2008-01-01

    Information about several past and present experimental and prototypical facilities based on High Temperature Gas-Cooled Reactor (HTGR) concepts have been examined to assess the potential of these facilities for use in this benchmarking effort. Both reactors and critical facilities applicable to pebble-bed type cores have been considered. Two facilities - HTR-PROTEUS of Switzerland and HTR-10 of China and one conceptual design from Germany - HTR-PAP20 - appear to have the greatest potential for use in benchmarking the codes. This study presents the benchmark analysis of these reactors technologies by using MCNP4C2 and MVP/GMVP Codes to support the evaluation and future development of HTGRs. The ultimate objective of this work is to identify and develop new capabilities needed to support Generation IV initiative. (author)

  3. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    International Nuclear Information System (INIS)

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

  4. Behavior of radioactive organic iodide in an atmosphere of High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Nakashima, Mikio; Sagawa, Chiaki; Masaki, Nobuyuki; Hirabayashi, Takakuni; Aratono, Yasuyuki

    1990-06-01

    Formation and decomposition behavior of radioactive organic iodide have been studied in an atmosphere of High Temperature Gas-cooled Reactor (High Temperature Engineering Test Reactor, HTTR). Na 125 I was chosen for radioactive iodine source instead of CsI diffusing from coated fuel particles. Na 125 I adsorbed on graphite was heated in pure He and He containing O 2 or H 2 O atmosphere. The results obtained are as follows. It was proved that organic iodide was formed with organic radicals released from graphite even in He atmosphere. Thus, the interchange rate of inorganic iodide with organic iodide was remarkably decreased with prolonged preheat-treatment period at 1000degC. Organic iodide formed was easily decomposed by its recirculation into hot reaction tube kept at 900degC. When organic iodide was passed through powdered graphite bed, more than 70% was decomposed at 90degC. Oxygen and water vapour intermixed in He suppressed the interchange rate of inorganic iodide with organic iodide. These results suggest that organic iodide rarely exists in the pressure vessel under normal operating condition of HTTR, and, under hypothetical accident condition of HTTR, organic iodide fraction never exceeds the value used for a safety assessment of light water reactor. (author)

  5. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Roberge, A.; Felten, P.; Bastien, D.

    1979-01-01

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. The insulation retained an acceptable degree of resiliency. However, some fiber damage was observed within both the high and low temperature insulation blankets. A thermal analysis was conducted to correlate the hot duct heat transfer results with those obtained from the analytical techniques used for the HTGR design using a computer thermal model representative of the duct and test setup. The thermal performance of the insulation, the temperature gradient through the structural components, the heating load to the cooling system and the permeation flow effect on heat transfer were verified. Exellent correlation between the experimental data and the analytical techniques were obtained

  6. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  7. Simulation Of The Secondary Cooling System Failed For One Line Mode Of RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Susyadi; Sembiring, Tagor M; Isnaeni, Darwis

    2003-01-01

    Recently, an assessment of 15 MW power reactor RSG-GAS operated using one line cooling mode is under carried out, in which is in the same manner as BA TAN policy. At the power above mentioned, requirement for the research as well as isotop production has been fulfilled. To obtain the transient condition of 1 line-cooling mode, the simulation using RELAP5.MOD3.2 code was carried out. The simulation parameters interesting known are the inlet of primary coolant temperature after failed the secondary cooling system. At the first, reactor is operated at 15 MW steady state condition using 1 line-cooling mode. Primary coolant flow rate of 430 kg/s and secondary of 550 kg/s respectively. After that the decreasing is occurred due to stop of secondary cooling pump. Therefore the primary cooling inlet temperature to the core increase cause scram reactor by inserted control rod. During the transient occur, the characteristic of primary cooling temperature pattern change were obtained. The simulation result shows that the temperature increase (ΔT) temperature to the reactor is 5,1 o C at the second of 85.5. Here is lower than ΔT for the two cooling mode of 10 o C. That temperature characteristic still tolerable against acceptable safety margin to the flow instability

  8. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

  9. Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

    International Nuclear Information System (INIS)

    Bryers, John; Ashmead, Simon

    2016-01-01

    EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

  10. Design of a spherical fuel element for a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W.F.G.; Kloosterman, J.L.; Van Dam, H.; Van der Hagen, T.H.J.J.

    2004-01-01

    A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k eff with burn-up. Nitride fuel ( 15 N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

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

    Science.gov (United States)

    Reihman, Thomas C.

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

  12. Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

    2009-10-15

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  13. Reliability Analysis of RSG-GAS Primary Cooling System to Support Aging Management Program

    Science.gov (United States)

    Deswandri; Subekti, M.; Sunaryo, Geni Rina

    2018-02-01

    Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed.One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

  14. Improving economics and safety of water cooled reactors. Proven means and new approaches

    International Nuclear Information System (INIS)

    2002-05-01

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

  15. Improving economics and safety of water cooled reactors. Proven means and new approaches

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-05-01

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

  16. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  17. Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1988-01-01

    Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

  18. Optimization of the steam generator project of a gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Sakai, Massao

    1978-01-01

    The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

  19. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  20. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    International Nuclear Information System (INIS)

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-01-01

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  1. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  2. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  3. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  4. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  5. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

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

  6. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

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

  7. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  8. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    Energy Technology Data Exchange (ETDEWEB)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the

  9. Full-fluence tests of experimental thermosetting fuel rods for the high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1981-01-01

    The irradiation performance of injected thermosetting fuel rods is compared to that of standard pitch-temperature gas-cooled reactor requirements. The primary objective of the experiments reported here was to obtain additional irradiation data at higher fluences for resin-based rods with intermediate binder char contents within the 15 to 30 wt% ''window of acceptability'' that had been previously established. 12 refs

  10. Technical evolution and operation of French CO2 cooled reactors (UNGG)

    International Nuclear Information System (INIS)

    Berthion, Y.

    1986-10-01

    The technical evolution of the five French CO 2 cooled reactors (UNGG) from 1981 to 1986 needs to be outlined. These technical evolutions concerned the fuel element of Bugey 1 which is now slightly enriched, as well as the load reduction operation required by the grid. In addition work in underway to increase the safety at the two St Laurent units, or to repair the hot steel upper-structures of Chinon-3 unit

  11. Thermocouple evaluation model and evaluation of chromel--alumel thermocouples for High-Temperature Gas-Cooled Reactor applications

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1977-03-01

    Factors affecting the performance and reliability of thermocouples for temperature measurements in High-Temperature Gas-Cooled Reactors are investigated. A model of an inhomogeneous thermocouple, associated experimental technique, and a method of predicting measurement errors are described. Error drifts for Type K materials are predicted and compared with published stability measurements. 60 references

  12. Application of the complex equilibrium code QUIL to cesium-impurity equilibria in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.D.; Lunsford, J.L.; Stark, W.A. Jr.

    1976-05-01

    An equilibrium analysis has been made of the fission-product cesium in the primary coolant loop of the high-temperature gas-cooled reactor (HTGR). The species distributions that result at equilibrium have been calculated for various conditions of reactor operation. The cesium species considered were the monomer, dimer, oxides, hydroxides, and the hydride. The effect of cesium sorption isotherms on graphite also was included in the analysis. During normal reactor operations, the abundant species of cesium were calculated to be elemental cesium, Cs, and the monomeric hydroxide, CsOH. Under most conditions of steam ingress, the abundant species was calculated to be CsOH. Cesium adsorbed onto graphite was stable under all steam-ingress conditions considered. Thermal transients above 1500 0 K were required for equilibrium transport of cesium from the core to the coolant. The analysis was carried out using the complex equilibrium code QUIL, designed and written with special emphasis on features that make it applicable to the fission-product problem

  13. Method of avoiding hazards resulting from accidents in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Dorner, S.; Schretzmann, K.; Schumacher, G.

    1984-01-01

    In water-cooled reactors, e.g. BWRs and PWRs, elemental hydrogen is released by hydrolysis (in-leakage). In case of an accident in these reactors or at emergency cooling of e.g., a gas-cooled reactor with water additional hydrogen is produced by chemical reactions of the water with the cladding material. In order to prevent hydrogen pressurizing and the formation of a detonating gas mixture, dry powder containers are provided for in the endangered compartments of the reactor. In case of danger powdered CuO, MnO 2 , Fe 2 O 3 , or CdO, the oxygen content of which recombines with the hydrogen, is ejected from them. In addition, an extinguishing substance with an anticatalytic resp. inhibition effect and/or an inert gas of the group N 2 , He, Ar, CO 2 may be admixed to the powder resp. powder mixture. (orig./PW)

  14. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  15. Draft pre-application safety evaluation report for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Williams, P.M.; King, T.L.; Wilson, J.N.

    1989-03-01

    This draft safety evaluation report (SER) presents the preliminary results of a pre-application design review for the standard modular high-temperature gas-cooled reactor (MHTGR) (Project 672). The MHTGR conceptual design was submitted by the U.S. Department of Energy (DOE) in accordance with the U.S. Nuclear Regulatory Commission(NRC) 'Statement of Policy for the Regulation of Advanced Nuclear Power Plants' (51 FR 24643), which provides for early Commission review and interaction. The standard MHTGR consists of four identical reactor modules, each with a thermal output of 350 MWt, coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe. The reactors are helium cooled and graphite moderated and utilize ceramically coated particle-type nuclear fuel. The design includes passive reactor-shutdown and decay-heat-removal features. The staff and its contractors at the Oak Ridge National Laboratory and the Brookhaven National Laboratory have reviewed this design with emphasis on those unique provisions in the design that accomplish the key safety functions of reactor shutdown, decay-heat removal, and containment of radioactive material. This report presents the NRC staff's technical evaluation of those features in the MHTGR design important to safety, including their proposed research and testing needs. In addition this report presents the criteria proposed by the NRC staff to judge the acceptability of the MHTGR design and, where possible, includes statements on the potential of the MHTGR to meet these criteria. However, it should be recognized that final conclusions in all matters discussed in this report require approval by the Commission. Final determination on the acceptability of the MHTGR standard design is contingent on receipt and evaluation of additional information requested from DOE pertaining to the adequacy of the containment design and on the following: (1) satisfactory resolution of open safety issues identified

  16. A fast converging CFD model for thermal hydraulic analysis of gas cooled reactor cores

    International Nuclear Information System (INIS)

    Chen, Gary; Anghaie, Samim

    1999-01-01

    A computational fluid dynamics (CFD) approach to the solution of Navier-Stokes equations for the thermal and flow fields of gas cooled reactor cores is presented. An implicit-explicit MacCormack method based on finite volume discretization scheme, in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve axisymmetric, thin-layer Navier-Stokes equations. This numerical method requires only the inversion of block bidiagonal systems rather than block tridiagonal systems, thus yielding savings in computer time and storage requirements. A two-layer algebraic eddy viscosity turbulence model is used in this study. The effects of turbulence are simulated in terms of the eddy viscosity coefficient, which is calculated for an inner and an outer region separately. An enthalpy-rebalancing scheme is implemented to allow the convergence solutions to be obtained with the application of a wall heat flux. The detailed computational analysis developed in this work is used to evaluate many different Nusselt number equations, property corrections, and axial distance corrections. The calculation based on this CFD model is compared with other published results. The good agreement indicates the usefulness of the presented model for the prediction of flow and temperature distributions for gas cooled reactor cores. (author)

  17. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    International Nuclear Information System (INIS)

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-01-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  18. Medium temperature carbon dioxide gas turbine reactor

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

    2004-01-01

    A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

  19. Study of the oxidation mechanisms between impurities and surfaces applied to the future gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Duval, A.

    2010-01-01

    Inconel 617, main candidate for the heat exchangers of the gas-cooled next generation of nuclear reactors has been investigated. Two different problems occurring in the cooling system splits the study into two parts. Oxidizing impurities contained in the coolant can cause severe corrosion at 850 C. Radioactive impurities, coming from the fission reaction of the core can, in another hand contaminate the cooling loop and cause radioprotection problem for the maintenance and dismantling operations. Firstly, oxidizing gas partial pressure influence on oxidation of IN 617 at 850 C was investigated varying oxygen and water vapour partial pressure between 1.10 -5 mbar and 200 mbar. Oxide layers were characterized using XPS, SEM, EDX, GD-OES, XRD. Influence of partial pressure on layers structure and composition was determined. Effect of water vapour and partial pressure on growth mechanisms were also investigated. The second part of this study is focused on diffusion of Ag, stable isotope of Ag-110m in IN617 alloy and in the oxide layer forming at its surface at 850 C. Concentration profiles were obtained by GD-OES calibrated analysis. Diffusion coefficient could be obtained from these diffusion profiles: volume diffusion and grain boundary diffusion coefficients for the diffusion in the alloy, and an apparent diffusion coefficient for the diffusion in the oxide, due to the porosity of the structure. (author) [fr

  20. Stability of test environments for performance evaluation of materials for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Edgemon, G.L.; Wilson, D.F.; Bell, G.E.C.

    1993-01-01

    Stability of the primary helium-based coolant test gas for use in performance ests of materials for the Modular High-Temperature Gas-Cooled Reactor (MHTGR) was determined. Results of tests of the initial gas chemistry from General Atomics (GA) at elevated temperatures, and the associated results predicted by the SOLGASMIX trademark modelling package are presented. Results indicated that for this gas composition and at flow rates obtainable in the test loop, 466 ± 24C is the highest temperature that can be maintained without significantly altering the specified gas chemistry. Four additional gas chemistries were modelled using SOLGASMIX trademark

  1. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  2. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  3. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

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

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

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

  5. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  6. Conceptual design study for the enhanced gas cooled reactor (EGCR)

    International Nuclear Information System (INIS)

    Nakano, M.; Sadahiro, D.; Ozaki, H.; Bryant, S.D.; Cheyne, A.; Gilroy, J.E.; Hulme, G.; Lennox, T.A.; Sunderland, R.E.; Beaumont, H.M.; Kida, M.; Nomura, M.

    2001-01-01

    The preliminary concept of the carbon dioxide cooled fast reactor EGCR has been studied as a Generation IV system. EGCR with MOX fuel has a very good core performance, a breeding ratio over 1.2, a long operating cycle of 24 months, and a high burnup of 150 GWd/t. The plant system is based on the successful AGR experience but provides 3600 MWth. Enhanced passive safety features are provided and a debris tray included. Preliminary costing studies show that EGCR can be competitive to LWRs and can be constructed on a similar schedule. This EGCR concept also shows development potential. (author)

  7. Dynamic simulation for scram of high temperature gas-cooled reactor with indirect helium turbine cycle system

    International Nuclear Information System (INIS)

    Li Wenlong; Xie Heng

    2011-01-01

    A dynamic analysis code for this system was developed after the mathematical modeling and programming of important equipment of 10 MW High Temperature Gas Cooled Reactor Helium Turbine Power Generation (HTR-10GT), such as reactor core, heat exchanger and turbine-compressor system. A scram accident caused by a 0.1 $ reactivity injection at 5 second was simulated. The results show that the design emergency shutdown plan for this system is safe and reasonable and that the design of bypass valve has a large safety margin. (authors)

  8. Analysis of pressure drop accidents in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kameoka, Toshiyuki

    1980-01-01

    Research and development are carried out on various problems in order to realize a multi-purpose, high temperature gas-cooled experimental reactor by Japan Atomic Energy Research Institute and others. In the experimental reactor in consideration at present, it is planned to flow helium at 1000 deg C and 40 atm. For the purpose, high temperature heat insulation structures are designed and developed, which insulate heat on the internal surfaces of pressure vessels and pipings. Consideration must be given to these internal heat insulation structures about the various characteristics in the working environmental temperature and pressure conditions, the measures for preventing the by-pass flow due to the formation of gaps and the abnormal leak of heat through the natural convection in the heat insulators and others. In this paper, the experimental results on the rapid pressure reduction characteristics of ceramic fiber heat insulation structures are reported. The ceramic fiber heat insulation structures have the features such as the application to uneven surfaces and penetration parts, the prevention of by-pass flow, and very low permeability. The problem is the restoring force after the high temperature compression. The experiment on rapid pressure reduction due to the accidental release of gas and the results are reported. (Kako, I.)

  9. A comparative study on recycling spent fuels in gas-cooled fast reactors

    International Nuclear Information System (INIS)

    Choi, Hangbok; Baxter, Alan

    2010-01-01

    This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

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

    International Nuclear Information System (INIS)

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

    1975-12-01

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

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

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

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

  12. Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

    International Nuclear Information System (INIS)

    2007-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

  13. Conceptual design study of high temperature gas-cooled reactor for plutonium incineration

    International Nuclear Information System (INIS)

    Goto, Minoru

    2013-01-01

    JAEA has started a conceptual design study of a Pu burner HTGR, which is called CBHTR (Clean Burn High Temperature gas-cooled Reactor). CBHTR’s fuel is TRISO-coated fuel particle with PuO 2 -YSZ (Yttria- Stabilized Zirconia) kernel, which increase proliferation resistance, safety of geological disposal, and Pu incineration. CBHTR can decrease Puf ratio from 60% to 20% with 520 GWd/t. In the future, 15% of electricity capacity is employed by 7 of CBHTRs and 59 of U-HTRs. JAEA has a R and D plan of manufacturing technology of TRISO-coated fuel with PuO 2 -YSZ kernel

  14. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Jianghai Li

    2017-01-01

    Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

  15. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    EL Khatib, H.H.A.

    2013-01-01

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

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

    International Nuclear Information System (INIS)

    Luo, Chending; Zhao, Fuqiang; Zhang, Na

    2014-01-01

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

  17. Risk-informed design of a pebble bed gas reactor

    International Nuclear Information System (INIS)

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

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

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

    2006-01-01

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

  19. Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1976-01-01

    Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

  20. A preliminary neutronic evaluation of the high temperature gas-cooled test reactor HTR-10 using the scale 6.0 code

    International Nuclear Information System (INIS)

    Sousa, Romulo V.; Fortini, Angela; Pereira, Claubia; Carvalho, Fernando R. de; Oliveira, Arno H.

    2013-01-01

    The High Temperature Gas-cooled Test Reactor HTR-10 is a 10 MW modular pebble bed type reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles. This reactor was built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China, and its first criticality was attained on December 1, 2000. The main objectives of the HTR-10 are to verify and demonstrate the technical and safety features of the modular HTGR (High Temperature Gas-cooled Reactor) and to establish an experimental base for developing nuclear process heat applications. In this work, using the Standardized Computer Analysis for Licensing Evaluation (SCALE) 6.0, a nuclear code developed by Oak Ridge National Laboratory (ORNL), the HTR-10 first critical core is modeled by the DEN/UFMG. The K eff was obtained and compared with the reference value obtained by the Idaho National Laboratory. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel. (author)

  1. Design measures in evolutionary water cooled reactors to optimize for economic viability

    International Nuclear Information System (INIS)

    Oh, S.J.; Yu, S.K.W.; Appell, B.

    1999-01-01

    Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

  2. Method for operating nuclear reactor

    International Nuclear Information System (INIS)

    Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

    1978-01-01

    Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

  3. Definition of breeding gain for the closed fuel cycle and application to a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

    2006-01-01

    In this paper a definition is given for the Breeding Gain (BG) of a nuclear reactor, taking into account compositional changes of the fuel during irradiation, cool down and reprocessing. A definition is given for the reactivity weights required to calculate BG. To calculate the effects of changes in the initial fuel composition on BG, first order nuclide perturbation theory is used. The theory is applied to the fuel cycle of GFR600, a 600 MWth Generation IV Gas Cooled Fast Reactor. This reactor should have a closed fuel cycle, with a BG equal to zero, breeding just enough new fuel during irradiation to allow refueling by only adding fertile material. All Heavy Metal is recycled in the closed fuel cycle. The result is that a closed fuel cycle is possible if the reprocessing has low losses ( 238 U, 15% Pu, and low amounts of the Minor Actinides. (authors)

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

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

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

  6. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Roberge, A.; Felten, P.; Bastien, R.

    1979-01-01

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. Pre and post test metallurgical analyses were conducted on the Hastelloy-X structures and reference specimens. The results gave evidence of aging in the form of noticeable changes in room temperature tensile and reduction in area parameters. The Hastelloy-X welds exhibited greater changes in properties due to thermal aging. The antifriction coating (Cr 3 C 2 ) performed well without spallation or excessive wear. (orig.)

  7. Remote handling equipment for the decommissioning of the Windscale Advanced Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Barker, A.; Birss, I.R.; Fish, G.

    1984-01-01

    A decision to decommission the Windscale Advanced Gas Cooled Reactor was taken shortly after reactor shutdown in 1981. The fuel has now been discharged and the decommissioning programme will last about 10-12 years. The paper describes the programme and objectives and deals with methods of handling and disposing of the radioactive waste material. The main new facility required is a Waste Packaging Building adjacent to the existing reactor in which the waste boxes will be filled, active waste encapsulated in concrete and the boxes cleaned, swabbed and monitored to comply with IAEA transport regulations. The handling machine concept and features are described. The assaying and packaging of the waste material, the control of box movement and the process of concrete encapsulation is described. The paper concludes with a description of the development programme to support the Project. The tasks include a study of cutting techniques, production and control of dust and smoke, viewing and lighting methods, filtration, decontamination and fixing of contamination

  8. FACTORS INFLUENCING HUMAN RELIABILITY OF HIGH TEMPERATURE GAS COOLED REACTOR OPERATION

    Directory of Open Access Journals (Sweden)

    Sigit Santoso

    2016-10-01

    ABSTRAK Peran dan tindakan operator pada reaktor berpendingin gas akan berbeda dengan peran operator pada operasi tipe reaktor lain. Analisis unjuk kerja operator dan faktor yang berpengaruh dapat dilakukan secara komprehensif melalui analisis keandalan manusia(HRA. Melalui HRA dampak dari kesalahan manusia pada sistem maupun cara untuk mengurangi dampak dan frekuensi kesalahan dapat diketahui. Makalah membahas faktor yang berpengaruh pada tindakan operator, yaitu pada kejadian kecelakaan pendingin reaktor gas bersuhu tinggi-HTGR. Analisis untuk kualifikasi faktor pembentuk kinerja(PSF dilakukan berdasarkan kurva keandalan fungsi waktu, dan metode keandalan manusia yang dikembangkan berdasar pada aspek kognitif yaitu Cognitive Reliability and Error Analysis Method (CREAM. Hasil analisis berdasar kurva keandalan fungsi waktu menunjukkan komponen waktu berkontribusi positif pada peningkatan keandalan operator (PSF<1 pada kondisi semua fitur keselamatan berfungsi sesuai rancangan. Sedangkan pada metoda analisis dengan pendekatan kognitif CREAM diketahui selain faktor ketersediaan waktu, faktor pelatihan dan rancangan HMI juga berkontribusi meningkatkan keandalan operator. Faktor pembentuk kinerja keseluruhan diketahui sebesar 0,25 dengan faktor kontribusi positif dominan atau berpengaruh pada penurunan kesalahan manusia adalah ketersediaan waktu (PSF=0,01, dan faktor kontribusi negatif dominan adalah prosedur dan siklus kerja (PSF=5. Nilai PSF tersebut sebagai faktor pengali dalam perhitungan probabilitas kesalahan manusia. Analisis faktor pembentuk kinerja perlu dikembangkan pada skenario kejadian lain untuk selanjutnya digunakan untuk perhitungan dan analisis keandalan manusia yang komprehensif dan perancangan sistem interaksi manusia mesin di ruang kendali. Kata kunci: PSF, HTGR, operator, ruang kendali, keandalan manusia

  9. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  10. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

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

    International Nuclear Information System (INIS)

    Nickel, H.

    1977-04-01

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

  12. Current status and future development of modular high temperature gas cooled reactor technology

    International Nuclear Information System (INIS)

    2001-02-01

    associated with these R and D programmes. Also, support of specific HTGR related research projects is included in the European Union's Fifth Framework Program beginning in 2000. Further opportunities and capabilities of the HTGR in the development of co-generation and non-electric applications are presented in Chapter 7. Spent fuel disposal and decommissioning are key issues that are significantly influencing the future of nuclear power. Chapter 8 addresses the anticipated manner of handling these areas within the PBMR and GT-MHR. Also addressed are the activities associated with spent fuel disposal and decommissioning of HTGRs previously shut down. The development and commissioning of any new nuclear plant concept is subject to risks and challenges to its commercialization. This is also evident in the closed cycle gas turbine, particularly with regard to the design and development of the power conversion system (PCS). The GT-MHR and the PBMR (as well as many other designs under consideration) incorporate state-of-the-art components in their PCS that must operate safely and efficiently for this concept to succeed. These components include magnetic bearings on the rotating machines, large compact plate-fin recuperator modules and seals between PCS components that have size, orientation or environmental operating characteristics yet to be fully demonstrated and proven. These challenges to the commercialization of the GT-MHR and PBMR are discussed in Chapter 9. The IAEA is advised on its activities in development and application of gas cooled reactors by the IWGGCR which is a committee of leaders in national programmes in this technology. The IWGGCR meets periodically to serve as a global forum for information exchange and progress reports on the national programmes, to identify areas of collaboration and to advise the IAEA on its programme. Countries with representation in the IWGGCR include Austria, China, France, Germany, Indonesia, Italy, Japan, the Netherlands, Poland, the

  13. The role of the IAEA in gas-cooled reactor development and application

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, J.; Kupitz, J. [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power

    1995-03-01

    The IAEA`s activities in gas-cooled reactor development focus on the four technical areas which are predicted to provide advanced HTGRs with a high degree of safety, but which must be proven. These are: (a) the safe neutronic behaviour of the core, (b) reliance on ceramic coated fuel particles to retain fission products even under extreme conditions, (c) the ability to dissipate decay heat by natural heat transport mechanisms, and (d) the safe behaviour of the fuel and reactor core under chemical attack (air or water ingress). The first three are subjects of Coordinated Research Programmes (CRPs) and the last was recently addressed in an information exchange meeting. CRPs are 3 to 6 years in duration, and often involve experimental activities. CRPs allow a sharing of efforts on an international basis and benefit from the experience and expertise of researchers from the participating institutes. (J.P.N.).

  14. Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.; Felten, P.

    1979-01-01

    A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

  15. Gas turbine modular helium reactor in cogeneration

    International Nuclear Information System (INIS)

    Leon de los Santos, G.

    2009-10-01

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  16. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K. D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above

  17. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  18. The effect of water vapor in the reactor cavity in a MHTGR [Modular High Temperature Gas Cooled Reactor] on the radiation heat transfer

    International Nuclear Information System (INIS)

    Cappiello, M.W.

    1991-01-01

    Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs

  19. Increasing the maximum daily operation time of MNSR reactor by modifying its cooling system

    International Nuclear Information System (INIS)

    Khamis, I.; Hainoun, A.; Al Halbi, W.; Al Isa, S.

    2006-08-01

    thermal-hydraulic natural convection correlations have been formulated based on a thorough analysis and modeling of the MNSR reactor. The model considers detailed description of the thermal and hydraulic aspects of cooling in the core and vessel. In addition, determination of pressure drop was made through an elaborate balancing of the overall pressure drop in the core against the sum of all individual channel pressure drops employing an iterative scheme. Using this model, an accurate estimation of various timely core-averaged hydraulic parameters such as generated power, hydraulic diameters, flow cross area, ... etc. for each one of the ten-fuel circles in the core can be made. Furthermore, distribution of coolant and fuel temperatures, including maximum fuel temperature and its location in the core, can now be determined. Correlation among core-coolant average temperature, reactor power, and core-coolant inlet temperature, during both steady and transient cases, have been established and verified against experimental data. Simulating various operating condition of MNSR, good agreement is obtained for at different power levels. Various schemes of cooling have been investigated for the purpose of assessing potential benefits on the operational characteristics of the syrian MNSR reactor. A detailed thermal hydraulic model for the analysis of MNSR has been developed. The analysis shows that an auxiliary cooling system, for the reactor vessel or installed in the pool which surrounds the lower section of the reactor vessel, will significantly offset the consumption of excess reactivity due to the negative reactivity temperature coefficient. Hence, the maximum operating time of the reactor is extended. The model considers detailed description of the thermal and hydraulic aspects of cooling the core and its surrounding vessel. Natural convection correlations have been formulated based on a thorough analysis and modeling of the MNSR reactor. The suggested 'micro model

  20. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  1. Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

  2. Construction and performance tests of a secondary hydrogen gas cooling system

    International Nuclear Information System (INIS)

    Sanokawa, K.; Hishida, M.

    1980-01-01

    With the aim of a multi-purpose use of nuclear energy, such as direct steel-making, an experimental multi-purpose high-temperature gas-cooled reactor (VHTR) is now being developed by the Japan Atomic Energy Research Institute (JAERI). In order to simulate a heat exchanging system between the primary helium gas loop and the secondary reducing gas system of the VHTR, a hydrogen gas loop as a secondary cooling system of the existing helium gas loop was completed in 1977, and was successfully operated for over 2000 hours. The objectives of constructing the H 2 secondary loop were: (1) To get basic knowledge for designing, constructing and operating a high-temperature and high-pressure gas facility; (2) To perform the following tests: (a) hydrogen permeation at the He/H 2 heat exchanger (the surfaces of the heat exchanger tubes are coated by calorizing to reduce hydrogen permeation), (b) thermal performance tests of the He/H 2 heat exchanger and the H 2 /H 2 regenerative heat exchanger, (c) performance test of internal insulation, and (d) performance tests of the components such as a H 2 gas heater and gas purifiers. These tests were carried out at He gas temperature of approximately 1000 0 C, H 2 gas temperature of approximately 900 0 C and gas pressures of approximately 40 kg/cm 2 G, which are almost the same as the operating conditions of the VHTR

  3. The design status of the United States Department of Energy modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Mills, Raymond R. Jr.

    1990-01-01

    The U.S. Department of Energy's Modular High Temperature Gas Cooled Reactor (MHTGR) is being designed using a systems engineering approach referred to as the integrated approach. The top level requirement for the plant is that it provides safe, reliable, economical energy. The safety requirements are established by the U.S. Licensing Authorities, principally the Nuclear Regulatory Commission. The reliability and economic requirements associated with the top level functions have been established in close coordination and cooperation with the electrical utilities and other potential users, and the nuclear supply industry. The integrated approach uses functional analysis to define the functions and sub-functions for the plant and to identify quantitatively how the various functions must be fulfilled. The top four functions associated with the MHTGR are: maintain safe plant operation; maintain plant protection; maintain control of radionuclide release; maintain emergency preparedness. In addition to meeting all U.S. Regulatory Requirements this advanced reactor concept is being designed to meet the following requirements: do not require sheltering or evacuating of anyone outside the plant boundary of 425 meters as a result of normal or abnormal plant operation; do not require operator action in order to accomplish the above sheltering and evacuation objectives and the design must be insensitive to operator errors; utilize inherent characteristics of materials to develop passive safety features; provide very long times for corrective actions following the initiation of an abnormal event before plant damage would be incurred

  4. Detection of gas-permeable fuel particles for highl 7490 temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thiele, B.A.; Stinton, D.P.; Costanzo, D.A.

    1980-01-01

    Fuel for High-Temperature Gas-Cooled Reactors (HTGR) consists of uranium oxide-carbide and thoria microspheres coated with layers of pyrolytic carbon and silicon carbide. The pyrolytic carbon coatings must be gas-tight to perform properly during irradiation. Therefore, particles must be carefully characterized to determine the number of defective particles (ie bare kernels, and cracked or permeable coatings). Although techniques are available to determine the number of bare kernels or cracked coatings, no reliable technique has been available to measure coating permeability. This work describes a technique recently developed to determine whether coatings for a batch of particles are gas-tight or permeable. Although most of this study was performed on Biso-coated particles, the technique applies equally well to Triso-coated particles. About 150 randomly selected Biso-particle batches were studied in this work. These batches were first subjected to an 18-hr chlorination at 15000C, and the volatile thorium tetrachloride released through cracked or very permeable coatings was measured versus chlorination time. Chlorinated batches were also radiographed to detect any thorium that had migrated from the kernel into the coatings. From this work a technique was developed to determine coating permeability. This consists of an 18-hr chlorination of multiple samples without measurement of the heavy metal released. Each batch is then radiographed and the heavy metal diffusion within each particle is examined so it can be determined if a particle batch is permeable, slightly permeable, or gas-tight. (author)

  5. Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

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

    1985-11-01

    This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

  6. Multi-purpose nuclear heat source for advanced gas-cooled reactor plants

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1993-01-01

    Nuclear power has the potential to be the ultimate green technology in that it could eliminate the need for burning fossil fuels with their polluting combustion products and greenhouse gases. This view is shared by many technologists, but it may be a generation before the public becomes convinced, and that will involve overcoming many safety, institutional, financial, and technical impediments. This paper addresses only the latter topic; a major theme being that for nuclear power to truly be a green technology and significantly benefit society, it must meet the needs of the full energy spectrum. Specifically, it must satisfy energy needs beyond just the electricity generating sector by today's nuclear plants. By virtue of its high temperature capability, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is the only type of reactor that has the potential to meet the wide range of energy needs that will emerge in the future. This paper discusses the nuclear heat source that gives the MHTGR multi-purpose capability, which is recognized today, but will not be implemented until early in the next century

  7. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  8. Assessment of gas cooled fast reactor with indirect supercritical CO2 cycle

    International Nuclear Information System (INIS)

    Hejzlar, P.; Driscoll, M. J.; Dostal, V.; Dumaz, P.; Poullennec, G.; Alpy, N.

    2006-01-01

    Various indirect power cycle options for a helium cooled Gas cooled Fast Reactor (GFR) with particular focus on a supercritical CO 2 (SCO 2 ) indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The Balance Of Plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and SCO 2 recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of 550 .deg. C, (2) advanced design with turbine inlet temperature of 650 .deg. C and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect SCO 2 recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR 'proximate-containment' and the BOP for the SCO 2 cycle is very compact. Both these factors will lead to reduced capital cost

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

    Science.gov (United States)

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

    2013-09-03

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

  10. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  11. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  12. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  13. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  14. HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

    International Nuclear Information System (INIS)

    Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.

    1976-05-01

    A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes

  15. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

  16. Perspectives on Understanding and Verifying the Safety Terrain of Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Carlson, Donald E.

    2014-01-01

    The inherent safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving coolant depressurization, air ingress, moisture ingress, and reactivity insertion. In addition to discussing associated uncertainties and potential measures to address them, the paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  18. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  19. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  20. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    International Nuclear Information System (INIS)

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  1. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

  2. Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

  3. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  4. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-07-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  5. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    International Nuclear Information System (INIS)

    1990-01-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  6. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  7. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  8. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

  9. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  10. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    Energy Technology Data Exchange (ETDEWEB)

    Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  11. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

  12. Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Rokhmadi

    2007-01-01

    Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

  13. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    Martin, R.; Roche, R.

    1964-01-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr

  14. Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

    International Nuclear Information System (INIS)

    Gokhan Yesilyurt; Hassan, Y.A.

    2003-01-01

    A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

  15. Eighth meeting of the International Working Group on Gas-Cooled Reactors, Vienna, 30 January - 1 February 1989. Summary report. Part 1

    International Nuclear Information System (INIS)

    1989-12-01

    The Eighth Meeting of the IAEA International Working Group on Gas-Cooled Reactors was held in Vienna, Austria, from 30 January - 1 February, 1989. The Summary Report (Part I) contains the Minutes of the Meeting

  16. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  17. Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

    International Nuclear Information System (INIS)

    Minatsuki, Isao; Mizokami, Yorikata

    2007-01-01

    The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

  18. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wi S.; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2008-10-15

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO{sub 2}) driven Brayton cycle. According to their suggestion SCO{sub 2} is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider.

  19. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    International Nuclear Information System (INIS)

    Jeong, Wi S.; Suh, Kune Y.

    2008-01-01

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO 2 ) driven Brayton cycle. According to their suggestion SCO 2 is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider

  20. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a