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Sample records for htgr type reactors

  1. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

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

  2. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  4. The market for HTGR type reactors

    International Nuclear Information System (INIS)

    Roehler, E.

    1986-01-01

    High-temperature-reactors with pebble-bed-reactor cores as a progressive reactor line, have been developed by BBC/HRB the Federal Republic of Germany over a period of 27 years and will soon be mature to be introduced to the market. They represent an important innovation in the field of reactor engineering. Due to its high degree of applicability on the power and heat market and its high flexibility regarding the site and fuel cycle the HTR is extremely suitable for providing energy to consumers, especially in countries using nuclear energy supply for the first time. (orig.) [de

  5. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

    The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

  6. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  7. Status, results and usefulness of risk analyses for HTGR type reactors of different capacity accessory to planning

    International Nuclear Information System (INIS)

    Kroeger, W.; Mertens, J.

    1985-01-01

    As regards system-inherent risks, HTGR type reactors are evaluated with reference to the established light-water-moderated reactor types. Probabilistic HTGR risk analyses have shown modern HTGR systems to possess a balanced safety concept with a risk remaining distinctly below legally accepted values. Inversely, the development and optimization of the safety concepts have been (and are being) essentially co-determined by the probabilistic analyses, as it is technically sensible and economically necessary to render the specific safety-related HTGR properties eligible for licensing. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Werth, G.

    1980-01-01

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

  9. HTGR type reactors in West Germany. Realizations and prospects

    International Nuclear Information System (INIS)

    Dauenert, U.

    1978-01-01

    The framework within which the research studies on high temperature reactors have been pursued in West Germany since 1960 is recalled. The principles guiding the present policy of the country in this domain are given: choice of a single technical conception that be applied both to reactors generating electricity and reactors producing high temperature heat for industrial processes such as coal gasification; to group the technical and industrial potentials of West Germany in this domain; financial and technical participation of electricity producers in the expected realizations; international cooperation. In this technique, West Germany is at present among the most advanced nations with the realization of a prototype 300 MWe reactor, financed by the electricity producers and a contribution of government [fr

  10. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

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

  11. Granular effect on the effective cross sections in the HTGR type reactors

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de.

    1975-01-01

    Effective cross section of bars for HTGR is studied from the point of view of heterogeneity. Microscopical heterogeneity due to grains is represented by a self-shielding factor, which is well determined [pt

  12. High-temperature Gas Reactor (HTGR)

    Science.gov (United States)

    Abedi, Sajad

    2011-05-01

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

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

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  18. Strategy to support HTGR fuel for the 10 MW Indonesia’s experimental power reactor (RDE)

    International Nuclear Information System (INIS)

    Taswanda Taryo; Geni Rina Sunaryo; Ridwan; Meniek Rachmawati

    2018-01-01

    The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power(25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and road map for the preparation of the RDE fuel plant construction with the involvement of national stake holders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia. (author)

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

  20. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

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

    International Nuclear Information System (INIS)

    1980-08-01

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

  2. Effects of the HTGR-gas turbine on national reactor strategies

    International Nuclear Information System (INIS)

    Ligon, D.M.; Brogli, R.H.

    1979-11-01

    A specific role for the HTGR in a national energy strategy is examined. The issue is addressed in two ways. First, the role of the HTGR-GT Binary cycle plant is examined in a national energy strategy based on symbiosis between fast breeder and advanced converter reactors utilizing the thorium U233 fuel cycle. Second, the advantages of the HTGR-GT dry-cooled plant operating in arid regions is examined and compared with a dry-cooled LWR. An event tree analysis of potential benefits is applied

  3. ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

    International Nuclear Information System (INIS)

    Conklin, J.C.

    2001-01-01

    1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

  4. 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

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

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

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

  6. 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

  7. 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)

  8. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  9. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

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

    International Nuclear Information System (INIS)

    Dedy Priambodo; Erlan Dewita; Sudi Ariyanto

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Xing Huiping; Xue Song

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  13. HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

    International Nuclear Information System (INIS)

    Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin

    2011-01-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

  14. Small demonstration HTGR concept

    International Nuclear Information System (INIS)

    Kiryushin, A.I.

    1989-01-01

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

  15. Process heat utilization from HTGR type reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  17. Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

    1972-10-15

    The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

  18. Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

    2014-01-01

    A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

  19. GCRA perspective on the HTGR-GT plant configuration

    International Nuclear Information System (INIS)

    1979-06-01

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

  20. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-15

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

  1. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

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

  2. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

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

  3. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Baldwin, D.H.

    1984-08-01

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

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

    International Nuclear Information System (INIS)

    Ray, A.; Bowman, H.F.

    1978-01-01

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

  5. HTGR analytical methods and design verification

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  6. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    Energy Technology Data Exchange (ETDEWEB)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  8. HTGR safety research concerns at NRC

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1982-01-01

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

  9. 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.)

  10. Personnel radiation exposure in HTGR plants

    International Nuclear Information System (INIS)

    Su, S.; Engholm, B.A.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  12. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

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

  13. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  14. Results of Koo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, F; Brighenti, G.; Chiodi, P. L.; Ghilardotti, G.; Giuliani, C.

    1974-10-15

    This paper describes k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  15. Flow of catalyst particles in a flue gas desulfurization plant; mass transfer in the domain of a detached flow - two examples (desulfurization, HTGR type reactor) for the application of big computers solving technical problems

    International Nuclear Information System (INIS)

    Achenbach, E.

    1988-01-01

    The research work of the Institute for Reactor Components is mainly experimental in character. Where possible, the experiments are accompanied by numerical calculations. This has the advantage of rendering parameter studies faster and more economical than is the case with experiments, so that physical contexts can become more apparent. However, these calculations are no substitute for experiments. The application of numerical calculations in connection with experimental results can now be demonstrated with two examples. The examples have been selected with the aim of making the presentation of the results sufficiently interesting for all those participating at the colloquium. The theoretical and experimental results are presented in the form of short films. (orig.) [de

  16. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  17. HTGR development in the United States of America

    International Nuclear Information System (INIS)

    Fox, J.E.

    1991-01-01

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

  18. HTGR Cost Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

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

  19. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  20. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  1. Status of international HTGR development

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-08

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

  4. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  5. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  6. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-01

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

  8. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  9. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  10. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  12. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  13. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  14. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  15. Study of the HTGR fission product migration at the Osiris experimental reactor

    International Nuclear Information System (INIS)

    Homme, A. l'; Lucot, M.

    1977-01-01

    A program of study on accidents in HTR reactor operation is presented: blowdown of primary coolant circuit, water inlet into the primary circuit, fuel element overheating by pipe logging or loss of cooling. These studies will be made in Aida irradiation loop in the pool of the Osiris reactor [fr

  16. Operating history report for the Peach Bottom HTGR. Volume I. Reactor operating history

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Baldwin, N.L.; Tomlin, R.W.

    1976-01-01

    The operating history for the Peach Bottom-1 Reactor is presented for the years 1966 through 1975. Information concerning general chemistry data, general physics data, location of sensing elements in the primary helium circuit, and postirradiation examination and testing of reactor components is presented

  17. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  18. Progress report on the k{sub infinity} measurements of HTGR type lattices with the oscillation technique at zero reactivity; Rapport d'advancement sur les mesures de K {sub infinite} des reseaux du type HTGR avec la technique de l'oscillateur a'reactivite' nulle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1971-01-15

    The principles of measuring k {sub infinity} for a HTGR lattice using the oscillation technique with zero reactivity were already presented at the ''9th reactor physics meeting of countries participating in the Dragon project''. A brief summary of the essential characteristics of the experiment is followed by a status report on present work.

  19. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  1. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  2. FBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Tsugio.

    1986-01-01

    Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

  3. Loop type LMFBR reactor

    International Nuclear Information System (INIS)

    Ito, Hiroyuki

    1989-01-01

    In conventional FBR type reactors, primary coolants at high temperature uprise at a great flow rate and, due to the dynamic pressure thereof, the free surface is raised or sodium is partially jetted upwardly and then fallen again. Then, a wave killing plate comprising a buffer plate and a deflection plate is disposed to the liquid surface of coolants. Most of primary sodium uprising from the reactor core along the side of the upper mechanism during operation collide against the buffer plate of the wave killing plate to moderate the dynamic pressure and, further, disperse radially of the reactor vessel. On the other hand, primary sodium passing through flowing apertures collides against the deflection plate opposed to the flowing apertures to moderate the dynamic pressure, by which the force of raising the free surface is reduced. Thus, uprising and waving of the free surface can effectively be suppressed to reduce the incorporation of cover gases into the primary sodium, so that it is possible to prevent in injury of the recycling pump, abrupt increase of the reactor core reactivity and reduction of the heat efficiency of intermediate heat exchangers. (N.H.)

  4. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Nick Soelberg

    2010-08-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  5. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Soelberg, Nick

    2010-01-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR 'full recycle' service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the 'pebble bed' approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R and D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in 'limited separation' or 'minimum fuel treatment' separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  6. HTGR fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  7. Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

    International Nuclear Information System (INIS)

    Chow, J.G.Y.; Soo, P.; Epel, L.

    1978-01-01

    A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

  8. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  9. FBR type reactor

    International Nuclear Information System (INIS)

    Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

    1996-01-01

    In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

  10. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  11. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  12. HTGR accident and risk assessment

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. Analysis and primary design of the I and C system architecture for HTGR-10 reactor

    International Nuclear Information System (INIS)

    Yang Zijue; Zhao Guoji

    1993-01-01

    The consideration of making good use of the-state-of-the-arts technology in designing advanced Instrumentation and Control System architecture is discussed. A fully distributed and fully micro-computerized, local network based Instrumentation and Control System is designed for the HTR-10 reactor. The advantages of the system architecture include high reliability and availability, flexibility, economics, etc. It also fits for other production processes

  14. HTGR Application Economic Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

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

  15. Summary of foreign HTGR programs

    International Nuclear Information System (INIS)

    1980-06-01

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

  16. Important viewpoints proposed for a safety approach of HTGR reactors in Europe

    International Nuclear Information System (INIS)

    Brinkmann, G.; Pirson, J.; Ehster, S.; Dominguez, M.T.; Mansani, L.; Coe, I.; Moormann, R.; Van der Mheen, W.

    2006-01-01

    The inherent safety features of modular High Temperature Reactors (HTRs) make events leading to severe core damage highly unlikely and constitute the main differentiating aspects compared to LWRs. Furthermore, while a known and stable regulatory environment has long been established for Light Water Reactors (LWRs), different ways of thinking may help to develop a more appropriate licensing process for HTR-based power plants. The HTR-L project funded by the European Commission in the 5th Framework Programme was dedicated to the definition of a common and coherent European safety approach and the identification of the main licensing issues for the licensing framework of the modular HTRs. Several topics were developed during the course of this project. Due to the characteristics of the HTR design, it has been necessary to define specific defence-in-depth requirements which have then been evaluated for implementation in the safety approach. Safety-related functions appropriate for the HTR design have also had to be identified and listed. On one hand, the different possible solicitations of the fuel particles constituted the starting point for the identification of the accidental conditions (by means of the Master Logic Diagrams methodology); these accidental conditions were classified and the most appropriate methods to consider ultra low probability severe accidents were examined. On the other hand, the elements constituting the source term and, in particular, the requirements for the confinement of radioactive products and the conditions required to prevent the need for a 'conventional' containment structure have been discussed. In the definition of the safety approach, attention has been paid to the need to maintain the potentially interesting economic perspectives of HTR reactors. Key issues to be addressed in the licensing process of the HTRs have also been identified. An innovative systems, structures and components classification method has been developed and

  17. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  18. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  19. Tank type nuclear reactors

    International Nuclear Information System (INIS)

    Naito, Kesahiro; Shimoyashiki, Shigehiro; Yokota, Norikatsu; Takahashi, Kazuo.

    1985-01-01

    Purpose: To improve the seismic proofness and the radiation shielding of LMFBR type reactors by providing the reactor with a structure reduced in the size and the weight, excellent in satisfactory heat insulating property and having radioactive material capturing performance. Constitution: Two sheets of ceramic plate members (for instance, mullite, steatite, beryllium ceramics or the like) which can be fabricated into plate-like shape and have high heat insulating property are overlapped with each other, between which magnetic heat-insulating material with magnetizing magnetic ceramics (for example, Lisub(0.5)Fesub(2.5)O 4 , Ni-Fe 2 O 4 , Fe-Fe 2 O 4 ) are sandwiched and the whole assembly is covered with metal coating material (for example, stainless steels). The inside of the coating material is evacuated or filled with an inert gas with low heat-conductivity (argon) at a pressure less than 1 kg/cm 2 abs, considering that the temperature goes higher and the inner pressure increases upon operation. In this way, the size of the laminated structure can be reduced to about 1/7 of the conventional case. The magnetic heat insulating materials can capture the magnetic impurities in sodium. (Kawakami, Y.)

  20. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  1. Waste management considerations in HTGR recycle operations

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  2. HTGR Industrial Application Functional and Operational Requirements

    International Nuclear Information System (INIS)

    Demick, L.E.

    2010-01-01

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

  3. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  4. Dynamic response of a multielement HTGR core

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  5. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  6. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  7. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  8. Pool-type reactor

    International Nuclear Information System (INIS)

    Hopkins, S.R.

    1977-01-01

    This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

  9. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  10. FBR type reactor

    International Nuclear Information System (INIS)

    Yamaoka, Mitsuaki

    1988-01-01

    Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

  11. BWR type reactors

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka

    1983-01-01

    Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

  12. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  13. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  14. The prospects of HTGR in China

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. BWR type reactors

    International Nuclear Information System (INIS)

    Hayashi, Katsuhisa; Watanabe, Shigeru.

    1983-01-01

    Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

  16. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  17. 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.

  18. FBR type reactor

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Zuketen, Atsushi.

    1982-01-01

    Purpose: To flatten the power distribution of a reactor core and shorten the breeding time. Constitution: The reactor core comprises a core region having fission products, an outer blanket region surrounding the outer side of the core region and having fertile material and an inner blanket region disposed within the core region and having fertile material. The axial thickness of the inner blanket region is made greater at the central portion and smaller at the peripheral portion of the inner blanket region, and the outermost peripheral end at the peripheral portion of the inner blanket region is opposed by way of the core region to the outer blanket region. In such an arrangement, the power decrease in the peripheral portion of the core region can be suppressed to thereby flatten the power distribution in the reactor core and shorten the breeding time. (Moriyama, K.)

  19. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  20. Natural convection type reactor

    International Nuclear Information System (INIS)

    Nakayama, Takafumi; Horiuchi, Tetsuo; Moriya, Kimiaki; Matsumoto, Masayoshi; Akita, Minoru.

    1988-01-01

    Purpose: To improve the reliability by decreasing the number of dynamic equipments and safely shutdown the reactor core upon occurrence of accidents. Constitution: A pressure relief valve and a pressurizing tank or gravitational water falling tank disposed to the main steam pipe of a reactor are installed in combination. Upon loss-of-coolant accident, the pressure relief valve is opened to reduce the pressure in the reactor pressure vessel to the operation pressure for each of the tanks, thereby enabling to inject water in the pressurizing tank at first and, thereafter, water in the gravitational water falling tank successively to the inside of the pressure vessel. By utilizing the natural force in this way, the reliability can be improved as compared with the case of pumped water injection. Further, by injecting an aqueous boric acid to a portion of a plurality of tanks, if the control rod insertion becomes impossible, aqueous boric acid can be injected. (Takahashi, M.)

  1. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  2. ORNL's NRC-sponsored HTGR safety and licensing analysis activities for Fort St. Vrain and advanced reactors

    International Nuclear Information System (INIS)

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

    1985-01-01

    The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. The program has also emphasized code verification, using Fort St. Vrain data where applicable, and comparing results with industry-generated codes. By the use of model and parameter adjustment routines, safety-significant uncertainties have been identified. A major part of the analysis work has been done for the Fort St. Vrain HTGR, and has included analyses of FSAR accident scenario re-evaluations, the core block oscillation problem, core support thermal stress questions, technical specification upgrade review, and TMI action plan applicability studies. The large, 2240-MW(t) cogeneration lead plant design was analyzed in a multi-laboratory cooperative effort to estimate fission product source terms from postulated severe accidents

  3. FBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Azekura, Kazuo; Inoue, Kotaro.

    1981-01-01

    Purpose: To decrease power fluctuations due to burning of blanket fuel element clusters by partially replacing the fertile materials in the blanket fuel element clusters with fissile materials. Constitution: Fertile materials in the radial blanket fuel element clusters disposed to the outside or inside of the reactor core are partially replaced with fissile materials. Since the power density of the fissile materials is at the maximum in the initial burning stage and decreases as the burning proceeds, the power density of the materials which is smaller in the initial burning stage and becomes greater with the burning by the neutron-accumulated plutonium is offset. Accordingly, the power fluctuations in the blanket fuel element clusters due to the burning made smaller thereby enable to form a reactor core with less power fluctuations due to burning under the constant coolant flow rate depending on the power in the final burning stage where the blanket power is maximum. (Moriyama, K.)

  4. FBR type reactors

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

    1986-01-01

    Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

  5. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  6. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  7. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  8. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  9. Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

    International Nuclear Information System (INIS)

    Von Lensa, W.

    1998-01-01

    Action on 'Safety-related Innovative Nuclear Reactor Technology Elements - R and D - (SINTER) Network' both aim at the identification of priority items for sustainable innovations of nuclear technologies and work-shared European collaboration structures. Such an approach can also be realised for future R and D on HTGR-related R and D under the umbrella of the IAEA as already proposed by the 'International Working Group on Gas-Cooled Reactors (IWGGCR)' in 1996 and illustrated in this paper for the construction of a 'Global HTGR R and D Network (GHTRN)'. 3 refs

  10. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  11. BWR type reactors

    International Nuclear Information System (INIS)

    Tsunoyama, Shigeaki; Tanabe, Akira.

    1979-01-01

    Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

  12. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part describes the MSBR core (data presented are from ORNL 4541). The principal characteristics of the core are presented in tables together with plane and elevation drawings, stress being put upon the reflector, and loading and unloading. Neutronic, and thermal and hydraulic characteristics (core and reflectors) are more detailed. The reasons why a graphite with a tight graphite layer has been chosen are briefly exposed. The physical properties of the standard graphite (irradiation behavior) have been determined for an isotropic graphite with fine granulometry; its dimensional variations largely ressemble that of Gilsonite. The mechanical stresses computed (Wigner effect) do not implicate in any way the graphite stack [fr

  13. LMFBR type reactors

    International Nuclear Information System (INIS)

    Sakurai, Akio; Matsushita, Kazuo.

    1985-01-01

    Purpose: To surely prevent the vibrational displacement of a rector core, and remove the sliding portions so as to avoid sticking or localized load concentration. Constitution: Cylindrical vibration-damping walls are protruded from the inner surface at the bottom of the main vessel. The vibration-damping walls constituted with outer and inner walls defining a coolant passage surround the reactor core support structures with small gaps between the inner surface of the inner wall and the outer circumferential surface of the core support structures. If the core support structures tend to displace horizontally due to earthquakes or the likes, the small gaps are varied and the coolants flow through the gap. Vibration-damping for the core support structures can be obtained by the flowing resistance to the coolants. (Yoshino, Y.)

  14. HTGR Measurements and Instrumentation Systems

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. Overview of gas-cooled reactor systems, their importance and their interactions

    International Nuclear Information System (INIS)

    Kasten, P.R.; Spiewak, I.; Tobias, M.L.

    1975-01-01

    The economic interactions between fueling, separative work, and capital requirements are illustrated for HTGR, GCFR, HTGR-GT, VHTR, LWRs and LMFBRs. The influence of finite low-cost uranium resources and of extensive LWR application within the next two decades on reactor use is also discussed. Technological developments required for the practical application of HTGRs, GCFRs, HTGR-GT and VHTRs are presented, along with the importance and environmental effects features of these applications. The technical advantages and disadvantages associated with use of the uranium and the thorium fuel cycles in HTGRs are given, including the implications a given fuel cycle has on fuel recycle and mined-fuel requirements. The influence of core design on HTGR fuel and coolant temperatures and on associated performance features are illustrated by considering prismatic and pebble-bed type cores. Finally, several scenarios relative to the development of the HTGR, GCFR, HTGR-GT and VHTR are presented. (auth)

  16. Recent evolution of HTGR instrumentation in the USA

    International Nuclear Information System (INIS)

    Rodriguez, C.

    1982-06-01

    The reactor instrumentation system for the 2240 MW(t) HTGR includes ex-core neutron detectors for automatic nuclear power control, separate ex-core neutron detectors for automatic protection purposes (reactor trip), reactor core outlet thermocouples that measure the temperature of the primary coolant (helium) as it exits the nuclear core, cold helium thermocouples that measure the temperature of the primary coolant as it enters the core, external pressure differential gages that measure primary coolant flow, in-core fission chambers that are utilized to map neutron flux, and ex-core primary coolant moisture monitors. All of these subsystems, except for the in-core flux mapping units, are also part of the Fort St. Vrain HTGR, which has provided significant experience for the design of the new system. In-core flux mapping is not necessary at FSV for normal operation because its relatively small core is fairly ''visible'' from the location of the ex-core instruments. However, temporary in-core fission couples, microphones, and displacement sensors, as well as sensitive ex-core accelerometers were utilized to identify periodic core block lateral movement and measure neutron flux and primary coolant temperatures. A search for in-core sensors to facilitate mapping neutron flux distributions in the larger core of the 2240 MW(t) HTGR has led to the selection of a high temperature fission chamber, which has been tested up to 1000 deg. C at General Atomic. The chamber shows adequate signal to noise ratio and repeatability. Other reactor instruments planned for the 2240 MW(t) are of the FSV type (i.e. thermocouples) or improved versions of the FSV design (i.e. moisture monitors). New concepts such as acoustic thermometers are also being considered

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki

    1997-11-01

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

  19. Progress of independent feasibility study for modular HTGR demonstration plant to be built in China

    International Nuclear Information System (INIS)

    He Jiachen

    1989-01-01

    Many regions in China are suffering from shortage of energy as a result of the rapid growth of the national economy, for example, the growth rate of national production in 1988 reached 11.2%. A great number of coal fired plants have been built in many industrial areas. However, the difficulties relating to the transportation of coal and environmental pollution have become more and more serious. The construction of hydropower plants is limited due to uneven geographic conditions and seasons. For these reasons China needs to develop nuclear power plants. Nowadays, it has been decided, that PWR will be the main reactor type in our country, but in some districts or under some conditions modular HTGR may have distinct advantages and become an attractive option. The possible plant site description and preliminary result of economic analysis of modular HTGR type reactor are briefly discussed in this presentation

  20. HTGR Application Economic Model Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-01-01

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

  1. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  2. Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

    International Nuclear Information System (INIS)

    Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

    1990-01-01

    In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

  3. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

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

    1979-04-01

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

  4. HTGR safety research program

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  5. Effect on non-linear soil-structure interaction due to base slab uplift on the seismic response of a high-temperature gas-cooled reactor (HTGR)

    International Nuclear Information System (INIS)

    Kennedy, R.P.; Short, S.A.

    1976-01-01

    In high seismic regions it has often been the practice to use oversized base slabs for the major nuclear power plant structures in order to prevent, or at least minimize the amount of dynamic base slab uplift which will result from the overturning moments developed during seismic ground motion. Two major reasons have been expressed as to why dynamic base slab uplift should be minimized: (1) As nuclear power plants are normally designed for seismic loadings based upon linear analysis, and since soil-structure interaction becomes nonlinear when only a portion of the base slab is in contact with the soil, linear elasticity analysis may be acceptable if base slab uplift occurs (as the resultant design loads may be incorrect), and (2) substantial uplift could cause excessive toe pressures in the supporting soil and significant impact forces when the slab recontacts the soil. The primary purpose of this paper is to evaluate the importance of the nonlinear soil-structure interaction effects resulting from substantial base slab uplift occurring during a seismic excitation. The structure for this investigation consisted of the containment building and prestressed reactor vessel (PCRV) for a typical HTGR plant. A simplified dynamic mathematical model was utilized consisting of a conventional lumped mass structure with soil-structure interaction accounted for by translational and rotational springs whose properties are determined by elastic half space theory. Three different site soil conditions (a rock site, a moderately stiff soil, and a soft soil) and two levels of horizontal ground motion (0.3 and 0.5 g earthquakes) were considered. (Auth.)

  6. HTGR steam generator development

    International Nuclear Information System (INIS)

    Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

    1976-01-01

    More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

  7. OMR type process heat reactor

    International Nuclear Information System (INIS)

    Franzetti, Franco.

    1974-01-01

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

  8. United Kingdom and USSR reactor types

    International Nuclear Information System (INIS)

    Lewins, Jeffery

    1988-01-01

    The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

  9. FY1983 HTGR summary level program plan

    International Nuclear Information System (INIS)

    1983-01-01

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

  10. Technical review of process heat applications using the HTGR

    International Nuclear Information System (INIS)

    Brierley, G.

    1976-06-01

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

  11. Exergy analysis of HTGR-GT

    International Nuclear Information System (INIS)

    Cao Jianhua; Wang Jie; Yang Xiaoyong; Yu Suyuan

    2005-01-01

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

  12. HTGR market assessment: interim report

    International Nuclear Information System (INIS)

    1979-09-01

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

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

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. 2000 MW(t) HTGR-DC-GT Modesto Site dry cooled model 346 concice

    International Nuclear Information System (INIS)

    1979-07-01

    Construction information is presented for a 800 MW(e) HTGR power reactor. The information is itemized for each reactor component or system and incudes quantity, labor hours, labor cost, material cost, and total costs

  16. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Higuchi, Shin-ichi; Kawashima, Masatoshi

    1987-01-01

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  17. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    Mann, L.G.

    1977-01-01

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  18. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

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

  19. Regulatory Framework of Safety for HTGR

    International Nuclear Information System (INIS)

    Huh, Chang Wook; Suh, Nam Duk

    2011-01-01

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

  20. High-temperature process heat applications with an HTGR

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  1. Present status of HTGR research and development, 1995

    International Nuclear Information System (INIS)

    1996-02-01

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

  2. HTGR experience, programs, and future applications

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  3. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  4. Friction, adhesion and corrosion performance of metallurgical coatings in HTGR-helium

    International Nuclear Information System (INIS)

    Engel, R.; Kleemann, W.

    1981-01-01

    The friction-, adhesion-, thermal cycling- and corrosion performance of several metallurgical coating systems have been tested in a simulated HTGR-test atmosphere at elevated temperatures. The coatings were applied to a solid solution strengthened Ni-based superalloy. Component design requires coatings for the protection of mating surfaces, since under reactor operating conditions, contacting surfaces of metallic components under high pressures are prone to friction and wear damage. The coatings will have to protect the metal surface for 30 years up to 950 0 C in HTGR-helium. The materials tested were various refractory carbides with or without metallic binders and intermetallic compounds. The coatings evaluated were applied by plasma spraying-, detonation gun- and chemical vapor deposition techniques. These yielded two types of coatings which employ different mechanisms to improve the tribiological properties and maintain coating integrity. (Auth.)

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

    International Nuclear Information System (INIS)

    1979-12-01

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

  6. HTGR gas turbine power plant preliminary design

    International Nuclear Information System (INIS)

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

    1973-01-01

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

  7. Approach on a global HTGR R and D network

    International Nuclear Information System (INIS)

    Lensa, W. von

    1997-01-01

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

  8. National HTGR safety program

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  9. Status of reprocessing technology in the HTGR fuel cycle

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  11. Steam generator design considerations for modular HTGR plant

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Development of high-strength concrete mix designs in support of the prestressed concrete reactor vessel design for a HTGR steam cycle/cogeneration plant

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.

    1985-01-01

    Design optimization studies indicate that a significant reduction in the size of the PCRV for a 2240 MW(t) HTGR plant can be effected through utilization of high-strength concrete in conjunction with large capacity prestressing systems. A three-phase test program to develop and evaluate high-strength concretes (>63.4 MPa) is described. Results obtained under Phase I of the investigation related to materials selection-evaluation and mix design development are presented. 3 refs., 4 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. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  15. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

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

    2014-10-01

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

  16. 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

  17. Reactor types for the future

    International Nuclear Information System (INIS)

    Hall, A.C.

    1990-01-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  18. Reactor types for the future

    Energy Technology Data Exchange (ETDEWEB)

    Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

    1990-06-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  19. HTGR structural-materials efforts in the US

    International Nuclear Information System (INIS)

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

    1982-07-01

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

  20. Tribological study on machine elements of HTGR components

    International Nuclear Information System (INIS)

    Nemoto, M.; Asanabe, S.; Kawaguchi, K.; Ono, S.; Oyamada, T.

    1980-01-01

    There are some tribological features peculiar to machines used in a high-temperature gas-cooled reactor (HTGR) plant. In this kind of plant, water-lubricated bearing combined with the buffer gas sealing system and/or gas-lubricated bearings are often applied in order to prevent degrading of the purity of coolant helium gas. And, it is essential for the reliability and safety design of the sliding members in the HTGR to obtain fundamental data on their friction and wear in high-temperature helium atmosphere. In this paper, the results of tests on these bearings and sliding members are introduced, which are summarized as follows: (1) Water-lubricated shrouded step thrust bearing and buffer gas sealing system were tested separately under the conditions simulated to those of circulators used in commercial plants. The results showed that each elements satisfies the requirements. (2) A hydrostatically gas-lubricated, pivoted pad journal bearing with a moat-shaped rectangular groove is found to be promising for use as a high-load bearing, which is indispensable for the development of a large-type circulator. (3) Use of ceramic coating and carbon graphite materials is effective for the prevention of adhesive wear which is apt to occur in metal-to-metal combinations. (author)

  1. Tribological study on machine elements of HTGR components

    International Nuclear Information System (INIS)

    Nemoto, Masaaki; Ono, Shigeharu; Asanabe, Sadao; Kawaguchi, Katsuyuki; Oyamada, Tetsuya.

    1981-11-01

    There are some tribological features peculiar to machines used in a high-temperature gas-cooled reactor (HTGR) plant. In this kind of plant, water-lubricated bearing combined with the buffer gas sealing system and/or gas-lubricated bearings are often applied in order to prevent degrading of the purity of coolant helium gas. And, it is essential for the reliability and safety design of the sliding members in the HTGR to obtain fundamental data on their friction and wear in high-temperature helium atmosphere. In this paper, the results of tests on these bearings and sliding members are introduced, which are summarized as follows: (1) Water-lubricated shrouded step thrust bearing and buffer gas sealing system were tested separately under the condition simulated to those of circulators used in commercial plants. The results showed that each elements satisfies the requirements. (2) A hydrostatically gas-lubricated, pivoted pad journal bearing with a moat-shaped rectangular groove is found to be promising for use as a high-load bearing, which is indispensable for the development of a large-type circulator. (3) Use of ceramic coating and carbon graphite materials is effective for the prevention of adhesive wear which is apt to occur in metal-to-metal combinations. (author)

  2. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  3. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

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

  4. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  6. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  7. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.

    2008-01-01

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  8. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  9. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    Ignatiev, V.; Devell, L.

    1995-01-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  10. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-01

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

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

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Tachibana, Yukio; Sun Yuliang

    2009-07-01

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

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

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Wang Hong

    2010-07-01

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

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

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Hino, Ryutaro; Yu Suyuan

    2006-06-01

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

  15. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  16. Loop-type FBR reactor

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

    1996-01-01

    The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

  17. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  18. CONTEMPT-G computer program and its application to HTGR containments

    International Nuclear Information System (INIS)

    Macnab, D.I.

    1976-03-01

    The CONTEMPT-G computer program has been developed by General Atomic Company to simulate the temperature-pressure response of a containment atmosphere to postulated depressurization of High-Temperature Gas-Cooled Reactor (HTGR) primary or secondary coolant circuits. The mathematical models currently used in the code are described, and applications of the code in examples of the atmospheric response of a representative containment to a variety of postulated HTGR accident conditions are presented. In particular, maximum containment temperature and pressure, equilibrated long-term prestressed concrete reactor vessel and containment pressures, and peak containment conditions following steam pipe ruptures are examined for a representative 770-MW(e) HTGR

  19. Strategies of development of reactor types

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

  20. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    May, J.; Perry, L.

    1980-01-01

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

  2. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  3. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  4. Applications of high-strength concrete to the development of the prestressed concrete reactor vessel (PCRV) design for an HTGR-SC/C plant

    International Nuclear Information System (INIS)

    Naus, D.J.

    1984-01-01

    The PCRV research and development program at ORNL consists of generic studies to provide technical support for ongoing PCRV-related studies, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities under this program have concentrated on the development of high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C plant, and the testing of models to both evaluate the behavior of high-strength concretes (plain and fibrous) and to develop model testing techniques. A test program to develop and evaluate high-strength (greater than or equal to 63.4 MPa) concretes utilizing materials from four sources which are in close proximity to potential sites for an HTGR plant is currently under way. The program consists of three phases. Phase I involves an evaluation of the cement, fly ash, admixtures and aggregate materials relative to their capability to produce concretes having the desired strength properties. Phase II is concerned with the evaluation of the effects of elevated temperatures (less than or equal to 316 0 C) on the strength properties of mixes selected for detailed evaluation. Phase III involves a determination of the creep characteristics and thermal properties of the selected mixes. An overview of each of these phases is presented as well as results obtained to date under Phase I which is approximately 75% completed

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  7. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  8. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  9. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  10. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

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

    International Nuclear Information System (INIS)

    1977-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kikstra, J.F

    1997-06-01

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

  13. Radioactive waste management at WWER type reactors

    International Nuclear Information System (INIS)

    1993-05-01

    This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

  14. 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.)

  15. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

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

  16. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

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

  17. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  18. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  20. HTGR depressurization analysis

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

  2. Prospects of HTGR process heat application and role of HTTR

    International Nuclear Information System (INIS)

    Shiozawa, S.; Miyamoto, Y.

    2000-01-01

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

  3. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  4. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  5. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  6. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  7. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  8. 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

  9. Dynamics and control modeling of the closed-cycle gas turbine (GT-HTGR) power plant

    International Nuclear Information System (INIS)

    Bardia, A.

    1980-02-01

    The simulation if presented for the 800-MW(e) two-loop GT-HTGR plant design with the REALY2 transient analysis computer code, and the modeling of control strategies called for by the inherently unique operational requirements of a multiple loop GT-HTGR is described. Plant control of the GT-HTGR is constrained by the nature of its power conversion loops (PCLs) in which the core cooling flow and the turbine flow are directly related and thus changes in flow affect core cooling as well as turbine power. Additionally, the high thermal inertia of the reactor core precludes rapid changes in the temperature of the turbine inlet flow

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

    International Nuclear Information System (INIS)

    Green, R.C.

    1981-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.C.

    1981-09-01

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

  12. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    OpenAIRE

    Setiadipura, T; Irwanto, D; Zuhair, Zuhair

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor ...

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-15

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

  17. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Zhu Yongjun

    1989-01-01

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

  19. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  20. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

  1. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  2. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  3. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

    In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

  4. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

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

  6. Utilization of HTGR on active carbon recycling energy system

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  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. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  9. HTGR strategy for reduced proliferation potential

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  10. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  11. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  12. Status of CHAP: composite HTGR analysis program

    International Nuclear Information System (INIS)

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

    1975-12-01

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

  13. Comparisons among different development ways of advanced reactors in China

    International Nuclear Information System (INIS)

    Guo Xingqu; Lin Jianwen; Wang Ruoli

    1992-03-01

    For the development of nuclear energy in the 21st century, China will select a new type reactor to develop, which will have higher fuel efficiency, high safety and better economics. The selection is among the types of FBR (fast breeder reactor), HTGR (high temperature gas-cooled reactor) and FFHR (fusion-fission hybrid reactor). Since the evaluation of advanced reactors involves many uncertain factors and the difficulty of quantization, both the AHP (analytic hierarchy process) method and expert consultation are adopted. Four aspects are taken in the norm system of AHP, i.e. safety, maturity of technology, economy and appropriateness. By using questionnaire method to experts and studying related documents, five types of advanced reactor are selected, i.e. oxide fueled FBR, metal fueled FBR, uranium fueled HTGR, U-Th fueled HTGR and FFBR. Their evaluation parameters are a comprehensively assessed and sorted. About 130 experts and professors who have been working in the research institutes and government agencies of nuclear field are asked to give their comments on the development of advanced reactors. The response rate of questionnaires is 86%, and the data collected are processed by computers. From the evaluation result of AHP method and expert consultation of the fast breeder reactor, especially, the metal fueled FBR, should have the priority in nuclear energy development in the 21st century in China

  14. Management feature of transuranic for HTGR and LWR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Long-lived actinides from spent fuels can cause potential long-term environ- mental hazards. The generation and incineration of transuranic in different closed fuel cycles were studied. U and Pu were recycled from spent fuel in the 250 MW high-temperature gas-cooled reactor-pebble-bed-module (HTR-PM) U-Pu fuelled core, and then PuO 2 and MOX fuel elements were designed based on this recycled U and Pu. These fuel elements were used to build up a new PuO 2 or MOX fuelled core with the same geometry of the original reactor. Characteristics of transuranic incineration with HTGR open and closed fuel cycles were studied with VSOP code, and the corresponding results from the light water reactor were compared and analyzed. The transuranic generation with HTGR open fuel cycle is almost half of the corresponding result of the light water reactor. Thus, HTGR closed fuel cycles can effectively burn transuranic. (authors)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. USNRC HTGR safety research program overview

    International Nuclear Information System (INIS)

    Foulds, R.B.

    1982-01-01

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

  17. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  18. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  19. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  20. Reduced risk HTGR concept for industrial heat application

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. Evaluation of the significance of inverse oxidation for HTGR graphites

    International Nuclear Information System (INIS)

    Lee, B.S.; Heiser, J. III; Sastre, C.

    1983-01-01

    The inverse oxidation refers to a higher mass loss inside the graphite than the outside. In 1980, Wichner et al reported this phenomenon (referred to as inside/out corrosion) observed in some H451 graphites, and offered an explanation that a catalyst (almost certainly Fe) is activated by the progressively increasing reducing conditions found in the graphite interior. Recently, Morgan and Thomas (1982) investigated this phenomenon is PGX graphites, and agreed on the existing mechanism to explain this pheomenon. They also called for attention to the possibility that this phenomenon may occur under HTGR (High Temperature Gas-Cooled Reactor) operating conditions. The purpose of this paper is to confirm the above mentioned explanation for this phenomenon and to evaluate the significance of this effect for HTGR graphites under realistic reactor conditions

  2. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

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

    1982-05-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  4. 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)

  5. Selection of JAERI'S HTGR-GT concept

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. Scaling laws for HTGR core block seismic response

    International Nuclear Information System (INIS)

    Dove, R.C.

    1977-01-01

    This paper discusses the development of scaling laws, physical modeling, and seismic testing of a model designed to represent a High Temperature Gas-Cooled Reactor (HTGR) core consisting of graphite blocks. The establishment of the proper scale relationships for length, time, force, and other parameters is emphasized. Tests to select model materials and the appropriate scales are described. Preliminary results obtained from both model and prototype systems tested under simulated seismic vibration are presented

  7. HTGR safety research at the Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Stroh, K.R.; Anderson, C.A.; Kirk, W.L.

    1982-01-01

    This paper summarizes activities undertaken at the Los Alamos National Laboratory as part of the High-Temperature Gas-Cooled Reactor (HTGR) Safety Research Program sponsored by the US Nuclear Regulatory Commission. Technical accomplishments and analysis capabilities in six broad-based task areas are described. These tasks are: fission-product technology, primary-coolant impurities, structural investigations, safety instrumentation and control systems, accident delineation, and phenomena modeling and systems analysis

  8. Study of air ingress accident of an HTGR

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1995-01-01

    Inherent properties of high temperature gas cooled reactors (HTGR) facilitate the design of HTGRs with high degree of passive safety performances. In this context, it is very important to establish a design criteria for a passive safe function for the air ingress accident. However, it is absolutely necessary to investigate the air ingress behavior during the accident before exploring the design criteria. The present paper briefly describes major activities and results of the air ingress research in our laboratory. (author)

  9. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

    1980-05-01

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

  11. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  12. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  13. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  14. Creep-Rupture Properties and Corrosion Behaviour of 21/4 Cr-1 Mo Steel and Hastelloy X-Alloys in Simulated HTGR Environment

    DEFF Research Database (Denmark)

    Lystrup, Aage; Rittenhouse, P. L.; DiStefano, J. R.

    Hastelloy X and 2/sup 1///sub 4/ Cr-1 Mo steel are being considered as structural alloys for components of a High-Temperature Gas-Cooled Reactor (HTGR) system. Among other mechanical properties, the creep behavior of these materials in HTGR primary coolant helium must be established to form part...

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    OpenAIRE

    Sudarmono Sudarmono

    2015-01-01

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

  17. HTGR R and D programs

    International Nuclear Information System (INIS)

    Neylan, A.J.; Brisbois, J.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. Recent activities on the HTGR for its commercialization in the 21st century

    International Nuclear Information System (INIS)

    Minatsuki, I.; Uchida, S.; Nomura, S.; Yamada, S.

    1997-01-01

    Currently, the greatest concern about energy is the need to rapidly increase the energy supply, while also conserving energy reserves and protecting the worldwide environment in the coming century. Furthermore, the direct use of thermal energy from nuclear reactors is an effective way to widen the application of nuclear energy. From this standpoint, Mitsubishi Heavy Industries (MHI) has been continuing the various activities related to the High Temperature Gas Cooled Reactor (HTGR). At present, MHI is participating in the High Temperature Engineering Test Reactor (HTTR) project, which is under construction at Oarai promoted by the Japan Atomic Energy Research Institute, as the primary fabricator. Moreover MHI has been conducting research and development to investigate the feasibility of HTGR commercialization in future. In this paper, the results of various studies are summarized to introduce our HTGR activities

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  1. Technical description of other types of reactors

    International Nuclear Information System (INIS)

    Vollmer, H.

    1977-01-01

    The paper reviews the development of reactor systems other than LWR, i. e. gas cooled reactors, heavy water reactors and fast breeders. The specific features of these reactors are discussed. Technical details on plant design of the various systems will be given as well as the present status-of-the-art. (orig.) [de

  2. Public acceptance of HTGR technology - HTR2008-58218

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  3. The acoustic environment in large HTGR's

    International Nuclear Information System (INIS)

    Burton, T.E.

    1979-01-01

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

  4. HTGR-GT systems optimization studies

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  5. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  6. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

  7. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  8. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  9. Self operation type reactor control device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1990-01-01

    A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

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

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1980-04-01

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

  12. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  13. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  14. The commercial application prospect of HTGR plant in China

    International Nuclear Information System (INIS)

    Wang Yingsu

    2008-01-01

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

  15. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  16. Economical and engineering aspects of modular-type fast reactors

    International Nuclear Information System (INIS)

    Kirillov, E.V.; Demidova, L.S.

    1989-01-01

    Economical and engineering characteristics for SAFR and PRISM modular-type reactors are analyzed on the basis of foreign papers. Dependence of economical characteristics for SAFR modules on their output is shown. Cost of power generation for the NPPs with PRISM reactor, LWR reactor and for coal thermal power plant is presented

  17. A reactivity accidents simulation of the Fort Saint Vrain HTGR

    International Nuclear Information System (INIS)

    Fainer, Gerson

    1980-01-01

    A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

    2015-01-01

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  20. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    International Nuclear Information System (INIS)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies

  1. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Costs and the environmental impact of radioactive waste treatment in reprocessing high-temperature gas-cooled reactor fuel

    International Nuclear Information System (INIS)

    Davis, W. Jr.

    1976-01-01

    A cost-benefit analysis and an analysis of the reduction in population dose from the use of different decontamination equipment in the off-gas system of a model plant for processing spent fuel from HTGR type reactors are presented

  4. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

    1997-01-01

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  5. Simulation of a pool type research reactor

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  6. Experimental determination of the Koo fuel temperature coefficient for an HTGR lattice

    Energy Technology Data Exchange (ETDEWEB)

    Agostini, P.; Benedetti, F.; Brighenti, G.; Chiodi, P. L.; Dell' Oro, P.; Giuliani, C.; Tassan, S.

    1974-10-15

    This paper describes temperature-dependent k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  7. Graphite oxidation in HTGR atmosphere

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  8. 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

  9. Simplified analysis of trasients in pool type liquid metal reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1987-01-01

    The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

  10. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  11. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

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

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

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

  13. Aging of reactor vessels in LWR type reactors

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-01-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs

  14. Status of the HTGR development program in Japan

    International Nuclear Information System (INIS)

    Saito, S.

    1991-01-01

    According to the revision of the Long-Term Program for Development and Utilization of Nuclear Energy issued by the Japanese Atomic Energy Commission, High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, will be constructed by the Japan Atomic Energy Research Institute (JAERI) in order to establish and upgrade the technology basis for an HTGR, serving at the same time as a potential tool for new and innovative basic research. The budget for the construction of the HTTR was approved by the Government and JAERI is now proceeding with the construction design of the HTTR, focussing the first criticality in the end of FY 1995. In order to establish and upgrade HTGR technology basis systematically and efficiently, and also to carry out innovative basic research on high temperature technologies, Japan will perform necessary R and D mainly at JAERI, which is a leading organization of the R and D. In addition, in order to promote the R and D on HTGRs more efficiently, Japan will promote the existing international cooperation with the research organizations in foreign countries. (author). 5 figs, 3 tabs

  15. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  16. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    Shatoff, H.; Charman, C.M.

    1983-01-01

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  18. Dynamic behaviour of a CAREM type reactor

    International Nuclear Information System (INIS)

    Abbate, P.; Doval, A.

    1990-01-01

    As complement to CAREM reactor design studies, behaviour analysis were made in a non-stationary regime, with the aim of developing plant systems and determining process variables variation ranges, characteristic of normal operating conditions, specifying alarm values for different variables, as well as for operating policies. Transient accidental scenes analysis were made, concluding that reactor characteristics provide security, maintaining the core integrity. (Author) [es

  19. 1170-MW(t) HTGR-PS/C plant application study report: Geismar, Louisiana refinery/chemical complex application

    International Nuclear Information System (INIS)

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

    1981-05-01

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

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

    International Nuclear Information System (INIS)

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

    2007-03-01

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

  1. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

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

    International Nuclear Information System (INIS)

    1993-01-01

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

  3. Bearing compartment seal systems for turbomachinery in direct-cycle HTGR power plants

    International Nuclear Information System (INIS)

    Adams, R.G.; Boenig, F.H.; Pfeifer, G.D.

    1977-10-01

    The direct-cycle High-Temperature Gas-Cooled Reactor (HTGR) employs a closed gas-turbine cycle with the primary reactor coolant (helium) as the working fluid. Design studies on this type of plant, carried out since 1971, have demonstrated, among other points, the advantages of the integrated arrangement, in which power from the cycle is transmitted to the electric generators by turbomachines completely enclosed in the reactor pressure vessel. A result of this arrangement is that the bearings are entirely enclosed within the primary coolant system of the reactor. An important aspect of the design of the turbomachinery is its prevention or minimization of the ingress of lubricants into the primary coolant system and its prevention of ingress of primary coolant into the bearing compartments. The design studies, which included thorough conceptual designs of the turbomachinery with emphasis on bearings and seals and their support systems showed that total exclusion of lubricant requires extremely complex seals and seal support systems. The variation of system low-end pressure with control actuation and the requirement that the bearing cavity pressure follow these variations were proved to further complicate the service system. The tolerance of even relatively minute amounts of entering lubricant during control transients will allow considerable simplification. This paper discusses the above-mentioned problems and their solutions in tracing the design evolution of a satisfactory bearing-compartment seals and service system. The resulting system appears to be feasible on the basis of experience with industrial gas turbines

  4. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

    1977-04-01

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

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

    International Nuclear Information System (INIS)

    Clark, M.J.

    1979-03-01

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

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  8. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  9. HTGR-INTEGRATED COAL TO LIQUIDS PRODUCTION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Anastasia M Gandrik; Rick A Wood

    2010-10-01

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

  10. HTGR-Integrated Coal To Liquids Production Analysis

    International Nuclear Information System (INIS)

    Gandrik, Anastasia M.; Wood, Rick A.

    2010-01-01

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

  11. Design of the HTGR for process heat applications

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  12. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  14. Nuclear closed-cycle gas turbine (HTGR-GT): dry cooled commercial power plant studies

    International Nuclear Information System (INIS)

    McDonald, C.F.; Boland, C.R.

    1979-11-01

    Combining the modern and proven power conversion system of the closed-cycle gas turbine (CCGT) with an advanced high-temperature gas-cooled reactor (HTGR) results in a power plant well suited to projected utility needs into the 21st century. The gas turbine HTGR (HTGR-GT) power plant benefits are consistent with national energy goals, and the high power conversion efficiency potential satisfies increasingly important resource conservation demands. Established technology bases for the HTGR-GT are outlined, together with the extensive design and development program necessary to commercialize the nuclear CCGT plant for utility service in the 1990s. This paper outlines the most recent design studies by General Atomic for a dry-cooled commercial plant of 800 to 1200 MW(e) power, based on both non-intercooled and intercooled cycles, and discusses various primary system aspects. Details are given of the reactor turbine system (RTS) and on integrating the major power conversion components in the prestressed concrete reactor vessel

  15. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  16. Present Status of HTGR Utilization System Development in Japan

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiaki

    2000-01-01

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

  17. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    Chichindaev, D.A.

    2001-01-01

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-12-01

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

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

    International Nuclear Information System (INIS)

    1981-12-01

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

  20. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  1. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  2. Potential of the HTGR hydrogen cogeneration system in Japan

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Mouri, Tomoaki; Kunitomi, Kazuhiko

    2007-01-01

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

  3. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  4. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  5. Measurement of flow field in the pebble bed type high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Lee, Sa Ya; Lee, Jae Young

    2008-01-01

    In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gascooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method

  6. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  7. Self-operation type power control device for nuclear reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru.

    1993-01-01

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

  8. 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.)

  9. Status of a reformer design for a modular HTGR in an in-line configuration

    International Nuclear Information System (INIS)

    Gluck, R.; Whitling, W.H.; Lipps, A.J.

    1984-01-01

    For the past several years the General Electric Company has had the technical lead on advanced concept studies for the Modular High Temperature Gas Cooled Reactor (HTGR) programs sponsored by the United States Department of Energy. The focus of the Modular Reactor System (MRS) effort is the development of a generic nuclear heat source capable of supplying heat to either a steam generator/electric cycle or a high temperature steam /methane reforming cycle. Some early ground rules for this study were that the reactor be designed for 950 deg. C direct cycle reforming and that the core be of the prismatic type. Since the prismatic core required control rods near the center of the core, the vertical in-line concept was selected to promote natural circulation cooling of the core for all potential transients except the depressurized core heatup transient. Although the requirement for a prismatic core has been eliminated for recent cost reduction studies, the vertical in-line configuration has been retained for its potential as the lowest cost configuration. This paper presents the results of recent design and analytical studies conducted to evaluate the feasibility of using a steam/methane reformer in a Vertical In-Line (VIL) arrangement with the generic nuclear heat source

  10. Present activity of the feasibility study of HTGR-GT system

    International Nuclear Information System (INIS)

    Muto, Y.; Miyamoto, Y.; Shiozawa, S.

    2001-01-01

    In JAERI a feasibility study of the High Temperature Gas-cooled Reactor-Gas Turbine (HTGR-GT) system has been carried out since January, 1997 as an assigned work by the Science and Technology Agency. The study aims at obtaining a promising concept of HTGR-GT system that yields a high thermal efficiency and at the same time is economically competitive. Designs of a few candidate systems will be undertaken and their power generation costs will be evaluated in parallel with design works, some experimental works such as the fabrication of a plate-fin type heat exchanger core and material tests will be carried out. The study will be continued till 2000 fiscal year. In 1997 fiscal year, a preliminary design of a direct cycle plant of 600 MWt was developed. A reactor inlet gas temperature of 460 deg. C, a reactor outlet gas temperature of 850 deg. C and a helium gas pressure of 6MPa were selected. Some advanced technologies were adopted such as a monolithic fuel compact and a control rod sheath made of carbon/carbon composite material. They were very effective to enhance the heat transfer of fuel and to reduce the core bypass flow. As a result, a power density of 6MW/m 3 and the maximum burnup of 10 5 MWD/ton were achieved. A single-shaft horizontal turbomachine of 3600 rpm was selected to ease the mechanical design of the rotor supported by magnetic bearings. The turbine, two compressors, a generator and six units of intercooler were placed in a turbine vessel, Plate-fin type recuperator and precooler are installed in a vertical heat exchanger vessel. By this design, a net thermal efficiency of 45.7% is expected to be achieved. To develop a high performance plate-fin recuperator, a core model of W200 mm x L200 mm x H200 mm with small fin size of 1.15 mm height was fabricated and as a result of tests, leak tightness, component strength and bonding appearance were found to be satisfactory. In 1998 fiscal year, a design of a direct cycle plant of 300 MWt is undertaken. The

  11. GTOROTO: a simulation system for HTGR core seismic behavior

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Nakamura, Yasuhiro; Onuma, Yoshio

    1980-07-01

    One of the most important design of HTGR core is its aseismic structure. Therefore, it is necessary to predict the forces and motion of the core blocks. To meet the requirement, many efforts to develop analytical methods and computer programs are made. A graphic simulation system GTOROTO with a CRT graphic display and lightpen was developed to analyze the HTGR core behavior in seismic excitation. Feature of the GTOROTO are as follows: (1) Behavior of the block-type HTGR core during earthquake can be shown on the CRT-display. (2) Parameters of the computing scheme can be changed with the lightpen. (3) Routines of the computing scheme can be changed with the lightpen and an alteration switch. (4) Simulation pictures are shown automatically. Hardcopies are available by plotter in stopping the progress of simulation pictures. Graphic representation can be re-start with the predetermined program. (5) Graphic representation informations can be stored in assembly language on a disk for rapid representation. (6) A computer-generated cinema can be made by COM (Computer Output Microfilming) or filming directly the CRT pictures. These features in the GTOROTO are provided in on-line conversational mode. (author)

  12. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  13. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  14. Conceptual designs of power tokamak-type thermonuclear reactors

    International Nuclear Information System (INIS)

    Shejndlin, A.E.; Nedospasov, A.V.

    1978-01-01

    Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  16. Supplement to the summary report of the specialists meeting on vessel concepts for gas-cooled reactors held in Lausanne, Switzerland, 23-25 October 1978

    International Nuclear Information System (INIS)

    1979-09-01

    The development of a reliable thermal insulation is an important problem common to all types of HTGR reactors. The French Commissariat a l'Energie Atomique and General Atomic have been pursuing such an aim for a steam cycle HTGR. It has been recognized very early in the design work that full size testing of the thermal barrier would be needed to prove its ability to withstand the normal and abnormal operational conditions. Two different sets of operational conditions tests have been defined: a) Long term test at nominal temperature conditions; b) Short term tests in abnormal temperature conditions. This paper is devoted to the second set of conditions

  17. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  18. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  19. European research and development on HTGR process heat applications

    International Nuclear Information System (INIS)

    Verfondern, Karl; Lensa, Werner von

    2003-01-01

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

  20. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  1. Thermal shielding device in LMFBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Hiroshi.

    1985-01-01

    Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

  2. Self operation type reactor scram device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1992-01-01

    A control rod having neutron absorbers therein is held by a curie point electromagnet by way of a control rod extension shaft. The electromagnet is suspended from a vertically movable driving shaft in an upper guide tube. Then, a heater is disposed at the lower portion in the inner side of the upper guide tube. Upon a function confirmation test, the electromagnet is at first pulled up to the inside of the upper guide tube. Subsequently, the electromagnet is heated by the heater by a temperature higher than the curie point of the temperature sensing magnetic material. If the function is normal, armature connected to the control rod extension tube is separated. With such a constitution, the electromagnetic portion is isolated from a coolant main stream, thereby enabling to avoid the cooling effect by the stream of coolants. Accordingly, the operation test for confirming the integrity of the function of the curie point electromagnet can be conducted while placing the electromagnet in the reactor core as it is during actual reactor operation. (I.N.)

  3. BWR type reactor and its operating method

    International Nuclear Information System (INIS)

    Ootsuji, Niro.

    1983-01-01

    Purpose: To regulate the control rod extraction operation such that an assumed control rod drop accident, if should occur, may not lead to further serious accidents, as well as enable to improve the working life of the control rod. Method: A plurality of control rods disposed among a plurality of fuel assemblies constituting the reactor core for suppressing the reactor core reactivity are divided into two groups depending on the descending speed, and the number of rods with a faster descending speed is set to less than 1/4 of the total number of the control rods. Then, the control rods are arranged such that those rods of the faster descending speed may be set every one another in any of the vertical, lateral and orthogonal directions. Further, it is always judged as to the possibility of extracting the control rods with the faster descending speed by a fast control rod extraction judging circuit to issue a signal to a control rod extraction inhibition circuit, so that the extraction operation for the control rods with the faster descending speed is started after all of the control rods with the slow descending speed have been extracted. Accordingly, if a control rod dropping accident should occur, abrupt power change can be avoided to thereby minimize the development of the accident. (Horiuchi, T.)

  4. Gas-cooled reactor programs: High-Temperature Gas-cooled Reactor Base-Technology Program. Annual progress report for period ending December 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.J.; Kasten, P.R.

    1979-06-01

    Progress in HTGR studies is reported in the following areas: fission product transport and coolant impurity effects, fueled graphite development, PCRV development, structural materials, characterization and standardization of graphite, and evaluation of the pebble-bed type HTGR.

  5. Generator technology for HTGR power plants

    International Nuclear Information System (INIS)

    Lomba, D.; Thiot, D.

    1997-01-01

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

  6. 60-MW/sub t/ methanation plant design for HTGR process heat

    International Nuclear Information System (INIS)

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

    1982-07-01

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

  7. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  8. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  9. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    International Nuclear Information System (INIS)

    Demick, L.E.

    2011-01-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  10. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

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

    International Nuclear Information System (INIS)

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

    2012-06-01

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

  14. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  15. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  16. Development of seismic analysis model for HTGR core on commercial FEM code

    International Nuclear Information System (INIS)

    Tsuji, Nobumasa; Ohashi, Kazutaka

    2015-01-01

    The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

  17. 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

  18. Derivation of criteria for primary circuit activity in an HTGR

    International Nuclear Information System (INIS)

    Su, S.D.; Barsell, A.W.

    1980-11-01

    This paper derives specific criteria for the circulating and plateout activity in the primary circuit for a 2170-MW(t) high temperature gas-cooled reactor-gas turbine (HTGR-GT) plant. Results show that for a design basis, (1) the circulating activity should be limited to 14,000 Ci Kr-88 (a principal nuclide) to meet both offsite dose and containment access constraint during normal operation and depressurization accidents, and (2) the plateout inventories for those important nuclides affecting shutdown maintenance should not exceed 10,000 Ci Ag-110m, 45,000 Ci Cs-134 and 130,000 Ci Cs-137. This paper presents bases and methodology for deriving such criteria and compares them with light water reactors. 5 tables

  19. Utilization of plutonium in HTGR and its actinide production

    International Nuclear Information System (INIS)

    Karin, S.; Brogli, R.; Lefler, W.; Nordheim, L.

    1976-01-01

    The HTGR is a potential plutonium consumer. In this function it would burn plutonium, produce electricity and the valuable fissile isotope U-233. The advantages of this concept are discussed but particular attention is given to the production and the destruction of the higher actinides due to the high burnup achievable in such a system. The presence of the strong resonances in the plutonium isotopes demanded an extension of the methods for evaluation of self-shielding factors, a different structure for broad groups, and the adaptation of the reactor codes to these changes. Specifications for coated plutonium particles were developed. Also procedures were determined to evaluate the alpha ray and neutron emission rates of the actinide nuclides. First cycle calculations were carried out to establish in detail the characteristics of the plutonium reactors and their results are given

  20. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  1. Double-walled tank type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishiguchi, Yohei.

    1993-01-01

    A secondary vessel containing a steam generator is disposed on a base slab, and a roof slab is disposed to the upper end opening of the base slab. A manometer sealing is formed between the upper end opening of the secondary vessel and the roof slab. A primary vessel is disposed in the second vessel for containing a reactor core therein. A communication pipeline system (equalizer) is disposed for communicating the cover gas space of the secondary vessel with the cover gas space of the primary vessel by way of the roof slab. The communication pipeline system comprises a breakable plate, a check valve which opens from the secondary system to the primary system, a closing valve and pipelines connecting them. Upon occurrence of a sodium-water reaction accident caused by rupture of heat transfer pipes of a steam generator in the secondary vessel, the breakable plate is broken to equalize the gas pressure by way of the communication pipelines. This can avoid external pressure buckling of the primary vessel. (I.N.)

  2. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    Concerning about global warming due to emission of greenhouse effect gas like CO 2 , it is essentially important to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of CO 2 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) which can supply high-temperature heat with high thermal efficiency as well as high heat-utilizing efficiency. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950 deg. C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). 2 refs, 2 figs, 1 tab., 2 photos

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  6. Long term review of research on light water reactor types

    International Nuclear Information System (INIS)

    Sumiya, Yutaka

    1982-01-01

    In Japan, 24 nuclear power plants of 17.18 million kWe capacity are in operation, and their rate of operation has shown the good result of more than 60% since 1980. One of the research on the development of light water reactors is the electric power common research, which was started in 1976, and 272 researches were carried out till 1982. It contributed to the counter-measures to stress corrosion cracking, thermal fatigue and the thinning of steam generator tubes, to the reduction of crud generation and the remote control and automation of inspection and maintenance, and to the verification of safety. The important items for the future are the cost down of nuclear power plant construction, the development of robots for nuclear power plants, the improvement of the ability to follow load variation, and the development of light water reactors of new types. It is necessary to diversify the types of reactors to avoid the effect of a serious trouble which may occur in one type of reactors. Tokyo Electric Power Co., Inc., thinks that the Japanese type PWRs having the technical features of KWU type PWRs are desirable for the future development. The compatibility with the condition of installation permission in Japan, the required design change and the economy of the standard design PWRs of KWU (1.3 million kW) have been studied since October, 1981, by KWU and three Japanese manufacturers. (Kako, I.)

  7. Two-dimensional vertical model seismic test and analysis for HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1983-02-01

    The resistance against earthquakes of high-temperature gas cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. In the paper the test results of seismic behavior of a half-scale two-dimensional vertical slice core model and analysis are presented. The following results were obtained: (1) With soft spring support of the fixed side reflector structure, the relative column displacement is larger than that for hand support but the impact reaction force is smaller. (2) In the case of hard spring support the dowel force is smaller than for soft support. (3) The relative column displacement is larger in the core center than at the periphery. The impact acceleration (force) in the center is smaller than at the periphery. (4) The relative column displacement and impact reaction force are smaller with the gas pressure simulation spring than without. (5) With decreasing gap width between the top blocks of columns, the relative column displacement and impact reaction force decrease. (6) The column damping ratio was estimated as 4 -- 10% of critical. (7) The maximum impact reaction force for random waves such as seismic was below 60% that for a sinusoidal wave. (8) Vibration behavior and impact response are in good agreement between test and analysis. (author)

  8. Two-dimensional horizontal model seismic test and analysis for HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1988-05-01

    The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)

  9. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  10. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  11. 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.

  12. 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

  13. 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

  14. Subharmonic excitation in an HTGR core

    International Nuclear Information System (INIS)

    Bezler, P.; Curreri, J.R.

    1977-01-01

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

  15. 1170-MW(t) HTGR-PS/C plant application study report: heavy oil recovery application

    International Nuclear Information System (INIS)

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

    1981-05-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Basic principles on the safety evaluation of the HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko

    2009-03-01

    As HTGR hydrogen production systems, such as HTTR-IS system or GTHTR300C currently being developed by Japan Atomic Energy Agency, consists of nuclear reactor and chemical plant, which are without a precedent in the world, safety design philosophy and regulatory framework should be newly developed. In this report, phenomena to be considered and events to be postulated in the safety evaluation of the HTGR hydrogen production systems were investigated and basic principles to establish acceptance criteria for the explosion and toxic gas release accidents were provided. Especially for the explosion accident, quantitative criteria to the reactor building are proposed with relating sample calculation results. It is necessary to treat abnormal events occurred in the hydrogen production system as an 'external events to the nuclear plant' in order to classify the hydrogen production system as no-nuclear facility' and basic policy to meet such requirement was also provided. (author)

  18. Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1980-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are being used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences

  19. Overview of HTGR heat utilization system development at JAERI

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    Verfondern, K.; Martin, R.C.; Moormann, R.

    1993-01-01

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

  1. The calculation - experimental investigations of the HTGR fuel element construction

    International Nuclear Information System (INIS)

    Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

    1985-01-01

    One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

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

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-04-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  5. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  6. New generation nuclear power units of PWR type integral reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

    1997-01-01

    Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

  7. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  8. Reactor design and safety approach for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Davies, S.M.; Yamaki, Hideo; Goodman, L.

    1984-06-01

    A tank type plant has been designed that offers compactness, high reliability under seismic and thermal transients, and a safety design approach that provides a balance between public safety and plant availability. This report provides a description of the design philosophy and safety features of the reactor

  9. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  10. 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)

  11. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1983-01-01

    This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

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

    International Nuclear Information System (INIS)

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

    2011-06-01

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

  13. 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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  15. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  16. Perspective channel-type reactor with enhanced safety

    International Nuclear Information System (INIS)

    Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

    1994-01-01

    Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

  17. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Sang-Guk

    2014-01-01

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

  18. Cesium transport data for HTGR systems

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  19. HTGR fuel particle crusher design evaluation

    International Nuclear Information System (INIS)

    Johanson, N.W.

    1978-10-01

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

  20. Sensitivity analysis of reflector types and impurities in 10 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Khamis, I.

    2007-01-01

    The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

  1. Sensitivity analysis of reflector types and impurities in a 10 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Khamis, I.

    2008-01-01

    The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

  2. 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)

  3. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

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

    International Nuclear Information System (INIS)

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

    2006-06-01

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

  5. Proven commercial reactor types: an introduction to their principal advantages and disadvantages

    International Nuclear Information System (INIS)

    Alesso, H.P.

    1981-01-01

    This study deals with the principal advantages and disadvantages of the five types of proven commercial reactors. A description of each class of commercial reactor (light water, gas-cooled, and heavy water) and their proven reactors is followed by a comparison of reactor types on the basis of technical merit, economics of operation, availability of technology, and associated political issues. (author)

  6. Overall aspects of control of ISIS-type nuclear reactor

    International Nuclear Information System (INIS)

    Amato, S.; Santinelli, A.

    1996-01-01

    The paper describes the main aspects related to the definition of main controls required to operate an ISIS-type nuclear power reactors. ISIS is a PWR-type intrinsically safe nuclear reactor designed by ANSALDO, based on density lock concept; it presents, between the other safety functions, self-depressurization and core cooling capability for unlimited time. Due to its specific characteristics, the ISIS reactor required to development of new control philosophy (if compared with actual nuclear power reactor) with the implementation of new control functions, for instance the density locks hot/cold interface locations control. This paper describes the main control functions implemented, their rationale, as well as the dynamic simulation performed to verify the adequacy of controls definitions. The dynamic simulations here described refers to a step-wise power ramp of 100-90-100 (% of nominal power) and to a power ramp of 100-50-100 with a slope of 5%/min; the results obtained have shown the ISIS capability to perform such operational transients, despite its innovative design was mainly focused on intrinsically safe behaviour. (author)

  7. Application of modern control theory to HTGR-plant

    International Nuclear Information System (INIS)

    Izaki, Makoto; Kubo, Hiroaki; Yamazaki, Eiji; Suzuki, Katsuo.

    1988-01-01

    The classical control theory approach to the multivariate control problem is to decouple the system intentionally and to treat each loop independently. As a result, final control system design is limited in complexity by the available mathematical techniques limitation and it's control performance is insufficient in many cases. The modern control theory approach based on the state variables to the problem provides far more powerful methods and more design flexibility than the classical control theory approach by the new mathematical formulation about the problem. The state variable feedback in formulating as an optimal regulator is the most effective way to obtain the desired control performance. In this report, some results of optimal regulator application to High Temperature Gas Cooled Reactor (HTGR) are shown. (author)

  8. 131I release from a HTGR during the LOFC accident

    International Nuclear Information System (INIS)

    Foley, J.E.

    1975-03-01

    The time-dependent release of 131 I from both the core and the containment building of a high temperature gas-cooled (HTGR) reactor during the loss of forced coolant (LOFC) accident is studied. A simplified core release model is combined with a containment building release model so that the total amount of the isotope released to the environment can be calculated. The time-dependent release of 131 I from the core during the LOFC accident is primarily a function of the time-dependent core temperatures and the failed fuel release constants. The most important factor in calculating the amount of the isotope released to the environment is the total amount released into the containment building. (U.S.)

  9. HTGR programme in the United States of America

    International Nuclear Information System (INIS)

    Fox, J.E.

    1991-01-01

    The HTGR is being developed by the US Department of Energy within the Division of HTGRs is reported. Fuel design, development and demonstration activities are being conducted by General Atomics and Oak Ridge National Laboratory. During FY-1990 the US continued work in cooperative projects with the KFA-Forschungszentrum Juelich and the Japan Atomic Energy Research Institute on post irradiation examination of fuel capsules and continued the Fission Product Transport Test Program with the French Commissariat a l'Energie Atomique in the COMEDIE in-pile loop at the SILOE reactor at Grenoble. Other activities included installation of the high temperature core-conduction-cooldown test furnace at ORNL which will be used for testing of irradiated fuel compacts under accident conditions. Finally, the US fuel performance experts participated in the MHTGR Cost Reduction Study which is a major effort within the US commercial MHTGR program. 1 tab

  10. Chemical thermodynamics of iodine species in the HTGR fuel particle

    International Nuclear Information System (INIS)

    Lindemer, T.B.

    1982-09-01

    The iodine-containing species in an intact fuel particle in the high-temperature gas-cooled reactor (HTGR) have been calculated. Assumptions include: (1) attainment of chemical thermodynamic equilibrium among all species in the open porosity of the particle, primarily in the buffer layer; and (2) fission-product concentrations in proportion to their yields. The primary gaseous species is calculated to be cesium iodide; in carbide-containing fuels, gaseous barium iodide may exhibit equivalent pressures. The condensed iodine-containing phase is usually cesium iodide, but in carbide-containing fuels, barium iodide may be stable instead. Absorption of elemental iodine on the carbon in the particle appears to be less than or equal to 10 -4 μg I/g C. The fission-product-spectra excess of cesium over iodine would generally be adsorbed on the carbon, but may form Cs 2 MoO 4 under some circumstances

  11. Thermal cycle efficiency of the indirect combined HTGR-GT power generation system

    Energy Technology Data Exchange (ETDEWEB)

    Muto, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-02-01

    High thermal efficiency of 50% could be expected in a power generation system coupling a high temperature gas-cooled reactor(HTGR) with a closed cycle gas turbine(GT). There are three candidate systems such as a direct cycle(DC), an indirect cycle(ICD) and an indirect combined cycle(IDCC). The IDCC could solve many problems in both the DC and the IDC and consists of a primary circuit and a secondary circuit where a topping cycle is a Brayton cycle and a bottoming cycle is a steam cycle. In this report, the thermal cycle efficiency of the IDCC is examined regarding configurations of components and steam pressure. It has been shown that there are two types of configurations, that is, a perfect cascade type and a semi-cascade one and the latter can be further classified into Case A, Case B and Case C. The conditions achieving the maximum thermal cycle efficiency were revealed for these cases. In addition, the optimum system configurations were proposed considering the thermal cycle efficiency, safety and plant arrangement. (author).

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

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Nakamura, Masashi

    1999-09-01

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

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. 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

  15. 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.

  16. Raw materials problems in connection with fast breeder type reactors

    International Nuclear Information System (INIS)

    Hirsch, H.; Kreusch, J.

    1981-01-01

    The power supply by the FBR type reactors which depends upon the availability of essential raw materials such as Cr and Mo for structural and special steels is supposed to be less ensured than supply by fossil-fueled power plants. This contribution tries to verify this statement by means of estimates of the annual Cr and Mo demand, of the resources, production and consumption as well as by a study of the possibilities of recycling and substituting Cr and Mo. The only realistic alternative to the fast breeder type reactor is supposed to be a soft path of development according to the principle of decentralization, utilization of renewable energy sources regard to environmental protection, and use of less sophisticated technology. (DG) [de

  17. Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.

    1975-01-01

    In the FRG three basic fuel element designs for application in high temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margin and presents aspects of their future development potential for operation in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. Summarizing it can be said that all the fuel elements as well as their components have proved their aptitude for the dual cycle systems in numerous fuel element and particle performance tests. To adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity further testing is still necessary, however. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.) [de

  18. HTGR generic technology program plan (FY 80)

    International Nuclear Information System (INIS)

    1980-01-01

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

  19. Temperature fluctuation reducing device for FBR type reactor

    International Nuclear Information System (INIS)

    Ootsuka, Fumio; Shiratori, Fumihiro.

    1991-01-01

    In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

  20. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  1. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  2. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages

    International Nuclear Information System (INIS)

    Jurado P, M.; Martin del Campo M, C.

    2005-01-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  3. Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

    International Nuclear Information System (INIS)

    Borges, E.M.; Braz Filho, F.A.

    1993-01-01

    The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

  4. Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  5. Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy

    International Nuclear Information System (INIS)

    Fukaya, Y.; Goto, M.

    2017-01-01

    Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR

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

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1975-12-01

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

  7. Uranium loss from BISO-coated weak-acid-resin HTGR fuel

    International Nuclear Information System (INIS)

    Pearson, R.L.; Lindemer, T.B.

    1977-02-01

    Recycle fuel for the High-Temperature Gas-Cooled Reactor (HTGR) contains a weak-acid-resin (WAR) kernel, which consists of a mixture of UC 2 , UO 2 , and free carbon. At 1900 0 C, BISO-coated WAR UC 2 or UC 2 -UO 2 kernels lose a significant portion of their uranium in several hundred hours. The UC 2 decomposes and uranium diffuses through the pyrolytic coating. The rate of escape of the uranium is dependent on the temperature and the surface area of the UC 2 , but not on a temperature gradient. The apparent activation energy for uranium loss, ΔH, is approximately 90 kcal/mole. Calculations indicate that uranium loss from the kernel would be insignificant under conditions to be expected in an HTGR

  8. Application of the lines of protection concept to the HTGR-SC/C

    International Nuclear Information System (INIS)

    1981-09-01

    This study of the application of the line of protection (LOP) concept to high temperature gas-cooled reactors (HTGRs) was motivated by a desire to develop a simple and straightforward HTGR safety concept that embodies many of the more complicated and seemingly conflicting concepts facing nuclear industry safety today. These concepts include: (1) defense in depth; (2) design basis events; (3) core damage events (degraded cores); (4) probabilistic analysis and risk assessment; (5) numerical safety goals; and (6) plant investment protection. The LOP concept described herein attempts to incorporate many of the important principles of each into a cohesive framework which provides an overall logic, meaning, and direction for conducting HTGR design and research activities

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1980-11-01

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

  10. Summary of ORNL work on NRC-sponsored HTGR safety research, July 1974-September 1980

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Delene, J.G.; Harrington, R.M.; Hatta, M.; Hedrick, R.A.; Johnson, L.G.; Sanders, J.P.

    1982-03-01

    A summary is presented of the major accomplishments of the Oak Ridge National Laboratory (ORNL) research program on High-Temperature Gas-Cooled Reactor (HTGR) safety. This report is intended to help the nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR. The ORNL program includes extensive work on dynamics-related safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification. Cooperative efforts with other programs are also described. Suggestions for near-term and long-term research are presented

  11. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  12. Accident transient processes at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    1982-01-01

    Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

  13. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  14. Present status of HTGR research and development

    International Nuclear Information System (INIS)

    1991-04-01

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

  15. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    International Nuclear Information System (INIS)

    Dodd, D.H.; Hienen, J.F.A. van.

    1995-10-01

    This report presents the results of task B.3 of the 'Technology Assessment of the High Temperature Reactor' project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL)

  16. The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, D.H.; Hienen, J.F.A. van

    1995-10-01

    This report presents the results of task B.3 of the `Technology Assessment of the High Temperature Reactor` project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL).

  17. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  18. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Quality control procedures for HTGR fuel element components

    International Nuclear Information System (INIS)

    Delle, W.W.; Koizlik, K.; Luhleich, H.; Nickel, H.

    1976-08-01

    The growing use of nuclear reactors for the production of electric power throughout the world, and the consequent increase in the number of nuclear fuel manufacturers, is giving enhanced importance to the consideration of quality assurance in the production of nuclear fuels. The fuel is the place, where the radioactive fission products are produced in the reactor and, therefore, the integrity of the fuel is of utmost importance. The first and most fundamental means of insuring that integrity is through the exercise of properly designed quality assurance programmes during the manufacture of the fuel and other fuel element components. The International Atomic Energy Agency therefore conducted an International Seminar on Nuclear Fuel Quality Assurance in Oslo, Norway from 24 till 28 May, 1976. This KFA report contains a paper which was distributed preliminary during the seminar and - in the second part - the text of the oral presentation. The paper gives a summary of the procedures available in the present state for the production control of HTGR core materials and of the meaning of the particular properties for reactor operation. (orig./UA) [de

  3. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  4. Method of freezing type dismantling for wasted reactors

    International Nuclear Information System (INIS)

    Tatsumi, Toshiyuki.

    1985-01-01

    Purpose: To enable to operate a cutting device in the air by placing a working table on ice while utilizing the ice as radiation shielding materials thereby prevent the diffusion of air contaminations. Method: Upon dismantling a BWR type reactor, ice is packed into a reactor container and a pressure vessel and frozen state is maintained by cooling coils disposed to the outer circumference of the pressure vessel. Then, an airtight hood is covered over the pressure vessel and a working table is rotatably disposed therein. Upon working, when the upper layer ice is melted by a heat pump and discharged, the airtight hood is lowered to a predetermined level. After freezing the melted portion again at the lowered level, cutting work is conducted by an operator in the hood. The cut pieces are conveyed after hoisting the airtight hood by a crane. The pressure vessel is dismantled by repeating the foregoing procedures. In this way, cut pieces can be recovered without falling them to the reactor bottom as in the conventional work in water. In addition, since the procedures are conducted while covering the airtight hood, diffusion of air contaminations can be prevented. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    1980-05-01

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

  6. Saturated steam turbines for power reactors of WWER-type

    International Nuclear Information System (INIS)

    Czwiertnia, K.

    1978-01-01

    The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

  7. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

  8. High temperature reactor and helium turbine for naval propeller (Study of feasibility and performances of the system)

    International Nuclear Information System (INIS)

    Brisbois, J.; Malherbe, J.; Rastoin, J.; Courau; Metayer.

    1976-01-01

    The nuclear reactor HTGR can get an outlet helium temperature greater than 800 deg C. That gives the means to use an helium turbine in a direct cycle. This type of reactor has been studied for a supply-ship (25,000t, 85,000Cv) and feasibility of such a system can be proved without employing any unknown materials. Because the weakness of helium activity, only the core can be shielded. All the propeller system is inside a containment which has to stand with a small over pressure after a core depressurisation. An efficiency of 35% is realized in a compact set up. This nuclear propeller get a very long core life 860FPD with constant worth- and very flexible working conditions. The HTGR direct cycle make a naval propeller very attractive [fr

  9. Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

    International Nuclear Information System (INIS)

    Nietert, R.E.; Abdelk-Khalik, S.I.

    1982-01-01

    An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

  10. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  11. WWER type reactors used as multipurpose nuclear power sources

    International Nuclear Information System (INIS)

    Fiala, J.; Mulak, J.

    1976-01-01

    Safety aspects are assessed of the siting of nuclear power installations in the vicinity of large housing estates and in areas with a high population density, mainly the aspect of the liquidation of the consequences of the maximum credible accident, i.e., the transversal rupture of the primary coolant circuit. The application of WWER type reactors as multipurpose nuclear power sources in Czechoslovakia is justified. It is shown that such a multipurpose nuclear power source differs from a purely condensation nuclear power plant mainly in the design of the secondary stage. The possibilities of such projects are indicated with a view to power and heat operation. (F.M.)

  12. 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.

  13. 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

  14. 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

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

    International Nuclear Information System (INIS)

    Goto, Minoru

    2015-03-01

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

  16. Identification of key amino acid residues in the hTGR5-nomilin interaction and construction of its binding model.

    Science.gov (United States)

    Sasaki, Takashi; Mita, Moeko; Ikari, Naho; Kuboyama, Ayane; Hashimoto, Shuzo; Kaneko, Tatsuya; Ishiguro, Masaji; Shimizu, Makoto; Inoue, Jun; Sato, Ryuichiro

    2017-01-01

    TGR5, a member of the G protein-coupled receptor (GPCR) family, is activated by bile acids. Because TGR5 promotes energy expenditure and improves glucose homeostasis, it is recognized as a key target in treating metabolic diseases. We previously showed that nomilin, a citrus limonoid, activates TGR5 and confers anti-obesity and anti-hyperglycemic effects in mice. Information on the TGR5-nomilin interaction regarding molecular structure, however, has not been reported. In the present study, we found that human TGR5 (hTGR5) shows higher nomilin responsiveness than does mouse TGR5 (mTGR5). Using mouse-human chimeric TGR5, we also found that three amino acid residues (Q77ECL1, R80ECL1, and Y893.29) are important in the hTGR5-nomilin interaction. Based on these results, an hTGR5-nomilin binding model was constructed using in silico docking simulation, demonstrating that four hydrophilic hydrogen-bonding interactions occur between nomilin and hTGR5. The binding mode of hTGR5-nomilin is vastly different from those of other TGR5 agonists previously reported, suggesting that TGR5 forms various binding patterns depending on the type of agonist. Our study promotes a better understanding of the structure of TGR5, and it may be useful in developing and screening new TGR5 agonists.

  17. Safety criteria for advanced HTGR concepts

    International Nuclear Information System (INIS)

    Kroeger, W.

    1989-01-01

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

  18. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. The water desalination complex based on ABV-type reactor plant

    International Nuclear Information System (INIS)

    Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

    1997-01-01

    A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

  20. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

  1. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  4. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  5. Overall plant concept for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Yamaki, Hideo; Davies, S.M.; Goodman, L.

    1984-01-01

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

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

    International Nuclear Information System (INIS)

    Bende, E.E.

    1997-01-01

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

  7. 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)

  8. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Goto, Tadashi; Inoue, Kotaro; Yamakawa, Masanori; Ikeda, Takashi.

    1988-01-01

    Purpose: To promote more positive forcive circulation of primary circuit fluids thereby increase the heat removing amount. Constitution: The primary side of an electromagnetic flow coupler type heat exchanger is opened to the primary fluid of a reactor, while the secondary side is connected with the secondary circuit comprising an air cooler and an electromagnetic pump. Since the secondary circuit stands-by during normal operation, the electromagnetic flow coupler does not operate and does not generate force for flowing primary circuit fluid. If flow due to the external force to the primary circuit fluid should occur in the electromagnetic flow coupler type heat exchanger, an electromagnetic force tending to flow the secondary circuit fluid is exerted oppositely. However the coupler undergoes reaction inertia of the fluid or flowing resistance, to exert in the direction of suppressing the flow, thereby prevent the heat loss. (Yoshihara, H.)

  9. Feasibility of monitoring the strength of HTGR core support graphite: Part III

    International Nuclear Information System (INIS)

    Morgan, W.C.; Davis, T.J.; Thomas, M.T.

    1983-02-01

    Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR). The results reported herein pertain to the development of techniques for monitoring the core-support blocks; the PGX graphite used in these studies is the grade used for the core-support blocks of the Fort St. Vrain HTGR, and is coarser-grained than the grades used in our previous investigations. The through-transmission ultrasonic velocity technique, developed for monitoring strength of the core-support posts, is not suitable for use on the core-support blocks. Eddy-current and ultrasonic backscattering techniques have been shown to be capable of measuring the density-depth profile in oxidized PGX and, combined with a correlation of strength versus density, could yield an estimate of the strength-depth profile of in-service HTGR core support blocks. Correlations of strength versus density and other properties, and progress on the development of the eddy-current and ultrasonic backscattering techniques are reported

  10. Very small HTGR nuclear power plant concepts for special terrestrial applications

    International Nuclear Information System (INIS)

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

    1983-01-01

    The role of the very small nuclear power plant, of a few megawatts capacity, is perceived to be for special applications where an energy source as required but the following prevail: 1) no indigenous fossil fuel source, in long transport distances that add substantially to the cost of oil, coal in gas, and 3) secure long-term power production for defense applications with freedom from fuel supply lines. A small High Temperature Gas-Cooled reactor (HTGR) plant could provide the total energy needs for 1) a military installation, 2) an island base of strategic significance, 3) an industrial community or 4) an urban area. The small HTGR is regarded as a fixed-base installation (as opposed to a mobile system). All of the major components would be factory fabricated and transported to the site where emphasis would be placed on minimizing the construction time. The very small HTGR plant, currently in an early stage of design definition, has the potential for meeting the unique needs of the small energy user in both the military and private sectors. The plant may find acceptance for specialized applications in the industrialized nations and to meet the energy needs of developing nations. Emphasis in the design has been placed on safety, simplicity and compactness

  11. Evaluation of creep-fatigue/ environment interaction in Ni-base wrought alloys for HTGR application

    International Nuclear Information System (INIS)

    Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira

    1986-01-01

    High Temperature Gas-cooled Reactor (HTGR) systems should be designed based on the high temperature structural strength design procedures. On the development of design code, the determination of failure criteria under cyclic loading and severe environments is one of the most important items. By using the previous experimental data for Ni-base wrought alloys, Inconel 617 and Hastelloy XR, several evaluation methods for creep-fatigue interaction were examined for their capability to predict their cyclic loading behavior for HTGR application. At first, the strainrange partitioning method, the frequency modified damage function and the linear damage summation rule were discussed. However, these methods were not satisfactory with the above experimental results. Thus, in this paper, a new fracture criterion, which is a modification of the linear damage summation rule, is proposed based on the experimental data. In this criterion, fracture is considered to occur when the sum of the fatigue damage, which is the function of the applied cyclic strain magnitude, and the modified creep damage, which is the function of the applied cyclic stress magnitude (determined as time devided by cyclic creep rupture time reflecting difference of creep damages by tensile creep and compressive creep), reaches a constant value. This criterion was successfully applied to the life prediction of materials at HTGR temperatures. (author)

  12. Development of toroid-type HTS DC reactor series for HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2015-11-15

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  13. Development of toroid-type HTS DC reactor series for HVDC system

    International Nuclear Information System (INIS)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-01-01

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  14. Study on commercial HTGR hydrogen production system

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo

    2000-07-01

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

  15. Nondestructive evaluation of the oxidation and strength of the Fort Saint Vrain HTGR support block

    International Nuclear Information System (INIS)

    Tingey, G.L.; Posakony, G.J.; Morgan, W.C.; Prince, J.M.; Hill, R.W.; Lessor, D.L.

    1982-04-01

    Non-destructive detection of changes in the strength of graphite support structures in a HTGR appears to be feasible using sonic velocity measurements where access for through transmission is possible. Therefore, future HTGR designs should consider providing such access. Where access is not available, strength changes can be correlated with oxidation profiles in the support member. These oxidation profiles can be determined non-destructively by a combination of eddy current measurements to detect near surface oxidation and sonic backscattering measurements designed to determine oxidation in depth. The Fort Saint Vrain reactor provides an operating reactor to test the applicability of the eddy current and sonic backscattering techniques for determination of oxidation in a support block. Furthermore, such tests in Fort Saint Vrain will supply base line data which will be useful in assuring an adequate strength of the support structure for the lifetime of the reactor. Equipment is, therefore, being developed for tests to be conducted during the next major refueling of the reactor

  16. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10 25 neutrons/m 2 to 8.5 x 10 25 neutrons/m 2 . Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating

  17. HTGR power plant hot reheat steam pressure control system

    International Nuclear Information System (INIS)

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

    1975-01-01

    A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

  18. Thermo-economic performance of HTGR Brayton power cycles

    International Nuclear Information System (INIS)

    Linares, J. L.; Herranz, L. E.; Moratilla, B. Y.; Fernandez-Perez, A.

    2008-01-01

    High temperature reached in High and Very High Temperature Reactors (VHTRs) results in thermal efficiencies substantially higher than those of actual nuclear power plants. A number of studies mainly driven by achieving optimum thermal performance have explored several layout. However, economic assessments of cycle power configurations for innovative systems, although necessarily uncertain at this time, may bring valuable information in relative terms concerning power cycle optimization. This paper investigates the thermal and economic performance direct Brayton cycles. Based on the available parameters and settings of different designs of HTGR power plants (GTHTR-300 and PBMR) and using the first and second laws of thermodynamics, the effects of compressor inter-cooling and of the compressor-turbine arrangement (i.e., single vs. multiple axes) on thermal efficiency have been estimated. The economic analysis has been based on the El-Sayed methodology and on the indirect derivation of the reactor capital investment. The results of the study suggest that a 1-axis inter-cooled power cycle has a similar thermal performance to the 3-axes one (around 50%) and, what's more, it is substantially less taxed. A sensitivity study allowed assessing the potential impact of optimizing several variables on cycle performance. Further than that, the cycle components costs have been estimated and compared. (authors)

  19. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  20. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    International Nuclear Information System (INIS)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments