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Sample records for advanced high-temperature reactor

  1. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

  2. Advanced High Temperature Reactor Systems and Economic Analysis

    International Nuclear Information System (INIS)

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

  3. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

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

  4. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

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

    International Nuclear Information System (INIS)

    Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H2) may become a significant application. The technology to produce H2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

  6. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  7. Commercial scale performance predictions for high-temperature electrolysis plants coupled to three advanced reactor types

    International Nuclear Information System (INIS)

    This paper presents results of system analyses that have been developed to assess the hydrogen-production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor - power-cycle combinations: a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to-hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable hydrogen production rates with the high-temperature helium-cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

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

  9. Advanced pebble bed high temperature reactor with central graphite column for future applications

    International Nuclear Information System (INIS)

    Design evaluations of the advanced pebble bed high temperature reactor, AHTR, with central graphite column are given. This reactor, as a nuclear heat source, is suitable for coal refinement as well as for electricity generation with closed gas turbine primary helium circuit. With this design of the central graphite column, it is possible to limit the core temperatures under the required value of about 1600deg C in case of accident conditions, even with higher thermal power and higher core inlet and outlet temperatures. The designs of core internals are described. The after heat removal system is integrated in the prestressed concrete reactor pressure vessel, which is based on the principals of natural convection. Research work is being carried out, whereby the sphencal fuel elements are coated with a layer of silicon carbide, to improve the corrosion resistance as well as the effectiveness of the fission products barrier. (orig.)

  10. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    International Nuclear Information System (INIS)

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  11. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  12. Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

    2012-06-01

    The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

  13. Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

    International Nuclear Information System (INIS)

    Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

  14. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  15. High temperature gas reactor

    International Nuclear Information System (INIS)

    The present invention provides a reflector block structure of a high temperature gas reactor in which graphite blocks are not failed even a containing cylinder loaded to a fuel exchanger collides against to secured reflectors upon loading and withdrawing fuel constitutional elements. Namely, a protection plate made of a metal material such as stainless steel is covered on the secured reflector blocks disposed to the upper most step among secured graphite reflector blocks constituting the reactor core. In addition, positioning guide grooves are formed on the protection plate for guiding the containing cylinder loaded to the fuel exchanger to the column of the reactor core constitutional elements. With such a constitution, even if the containing cylinder of fuel exchanger is hoisted down and collided against the inner circumferential edge of the secured reflector blocks due to deviation of the position and the direction upon exchange of fuels, the reflector blocks are not failed since the above-mentioned portion is covered with the metal protection plate. In addition, the positioning guide grooves lead the fuel exchanger to a predetermined column correctly. (I.S.)

  16. Coupling of RMC and CFX for analysis of Pebble Bed-Advanced High Temperature Reactor core

    International Nuclear Information System (INIS)

    Highlights: ► The CFD code CFX is used for whole pebble bed reactor core calculation. ► The Monte Carlo Code RMC and CFX are used for the coupling of neutronics and T-H. ► Coupled calculations for steady-state problem can reach stable results. ► Increasing the number of neutron histories is effective to improve accuracy. - Abstract: This paper introduces a steady-state coupled calculation method using the Monte Carlo Code RMC (Reactor Monte Carlo) and the Computational Fluid Dynamic (CFD) code CFX for the analysis of a Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. The porous media model is used in CFX modeling to simulate the pebble bed structure in PB-AHTR. The CFX model has also been validated against the RELAP5-3D model developed in the previous research. The script language PERL is used as a development tool to manipulate and control the entire coupled calculation. This research gives the conclusion that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations. However, due to the statistical errors of Monte Carlo method, the fluctuation of results still occurs. For the purpose of improving the accuracy, the paper applies and discusses two methods, of which increasing the number of neutron histories is an effective method.

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

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

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

  18. Progress in the Development of the Modular Pebble-Bed Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    This review article summarizes recent progress by students and faculty at U.C. Berkeley working on the development of the Pebble-Bed Advanced High Temperature Reactor (PB-AHTR). The 410-MWe PBAHTR is a liquid salt cooled reactor that operates at near atmospheric pressure and high power density (20 to 30 MW/m3, compared to 4.8 MW/m3 for helium cooled reactors). Operating with a core inlet temperature of 600 deg. C and outlet temperature of 704 deg. C, the PB-AHTR uses well understood materials of construction including Alloy 800H with Hastelloy N cladding for the reactor vessel and primary loop components, and graphite for core and reflector structures. Recent work by the NE 170 senior design class has developed physical arrangements for the major reactor and power conversion components, along with the structural design for the reactor building and turbine hall featuring seismic base isolation, design for aircraft crash protection, shielding analysis, and design of a multiple-zone ventilation and containment system to provide effective control of radioactive and chemical contamination. The resulting total building volume is 260 m3/MWe, compared to 343 m3/MWe to 486 m3/MWe for current large (1150 to 1600 MWe) LWR designs. These results suggest the potential for significant reductions in construction time and cost. Neutronics studies have verified the capability to design the PB-AHTR with negative fuel and coolant temperature reactivity coefficients, for both LEU and deep-burn TRU fuels. Depletion analysis was also performed to identify optimal core designs to maximize fuel utilization. The additional moderation provided by the coolant simplifies design to achieve optimal moderation, and the spent fuel volume is approximately half that of helium cooled reactors. In collaboration with the Czech Nuclear Research Institute, initial zero-power critical tests were performed to validate PB-AHTR neutronics models. Liquid salts are unique among candidate reactor coolants due

  19. Advanced reactor water cleanup system with high-temperature electrophoresis demineralization process as alternative to ion-exchange resin process

    International Nuclear Information System (INIS)

    The ion-exchange resin process has been widely applied to reactor water cleanup systems to remove impurities from the water used in boiling water reactors (BWRs). Toshiba has developed a high-temperature electrophoresis demineralization process as an alternative to the ion-exchange resin process for an advanced reactor water cleanup system. Since the new process uses only inorganic materials, high-temperature and high-pressure water can be fed directly to the system. The new system was confirmed to remove ions with high efficiency in a performance test using high-temperature and high-pressure water simulating BWR water. The advanced reactor water cleanup system will be greatly simplified because heat exchangers and resin-handling equipment are not required. It will also be economical due to reductions in heat loss and resin waste. (author)

  20. Thermal-hydraulics numerical analyses of Pebble Bed Advanced High Temperature Reactor hot channel

    International Nuclear Information System (INIS)

    Background: The thermal hydraulics behavior of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) hot channel was studied. Purpose: We aim to analyze the thermal-hydraulics behavior of the PB-AHTR, such as pressure drop, temperature distribution of coolant and pebble bed as well as thermal removal capacity in the condition of loss of partial coolant. Methods: We used a modified FLUENT code which was coupled with a local non-equilibrium porous media model by introducing a User Defined Scalar (UDS) in the calculation domain of the reactor core and subjoining different resistance terms (Ergun and KTA) to calculate the temperature of coolant, solid phase of pebble bed and pebble center in the core. Results: Computational results showed that the resistance factor has great influence on pressure drop and velocity distribution, but less impact on the temperature of coolant, solid phase of pebble bed and pebble center. We also confirmed the heat removal capacity of the PB-AHTR in the condition of nominal and loss of partial coolant conditions. Conclusion: The numerical analyses results can provide a useful proposal to optimize the design of PB-AHTR. (authors)

  1. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  2. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390°C for KF ZrF4). The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses

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

    International Nuclear Information System (INIS)

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

  4. Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

    International Nuclear Information System (INIS)

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

  5. The high temperature reactor

    International Nuclear Information System (INIS)

    The outstanding characteristics of the HTR could not save this well-proved sort from an eventful fate. They did ensure, however, that despite all hindrances and delays, extensive experiences with the HTR are available today which form a broad fundament for this type which is regarded as very important for the future. The most important starting point for the future might be the fact that now all industrial groups interested in the HTR have gathered for a joint deliberation of the situation and decision making. The guiding lines seem to be drawn already: there are broad fields of application on the heat market, but power generation also offers incentives, connected with reasons of fuel-saving. All things considered, today's world energy situation represents the biggest challenge and the most powerful impetus to this reactor which possesses the important basic lines of the innovation required in energy converting processes. (orig.)

  6. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  7. MOTHER MK II: An advanced direct cycle high temperature gas reactor

    International Nuclear Information System (INIS)

    The MOTHER (MOdular Thermal HElium Reactor) power plant concepts employ high temperature gas reactors utilizing TRISO fuel, graphite moderator, and helium coolant, in combination with a direct Brayton cycle for electricity generation. The helium coolant from the reactor vessel passes through a Power Conversion Unit (PCU), which includes a turbine-generator, recuperator, precooler, intercooler and turbine-compressors, before being returned to the reactor vessel. The PCU substitutes for the reactor coolant system pumps and steam generators and most of the Balance Of Plant (BOP), including the steam turbines and condensers, employed by conventional nuclear power plants utilizing water cooled reactors. This provides a compact, efficient, and relatively simple plant configuration. The MOTHER MK I conceptual design, completed in the 1987 - 1989 time frame, was developed to economically meet the energy demands for extracting and processing heavy oil from the tar sands of western Canada. However, considerable effort was made to maximize the market potential beyond this application. Consistent with the remote and very high labour rate environment in the tar sands region, simplification of maintenance procedures and facilitation of 'change-out' in lieu of in situ repair was a design focus. MOTHER MK I had a thermal output of 288 MW and produced 120 MW electrical when operated in the electricity only production mode. An annular Prismatic reactor core was utilized, largely to minimize day-to-day operations activities. Key features of the power conversion system included two Power Conversion Units (144 MWth each), the horizontal orientation of all rotating machinery and major heat exchangers axes, high speed rotating machinery (17,030 rpm for the turbine-compressors and 10,200 rpm for the power turbine-generator), gas (helium) bearings for all rotating machinery, and solid state frequency conversion from 170 cps (at full power) to the grid frequency. Recognizing that the on

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

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

    International Nuclear Information System (INIS)

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

  10. Qualification of metallic materials for application in advanced high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    As in conventional high temperature technology, the qualification of metallic materials for high temperature reactor (HTR) applications is based on creep behavior, fatigue properties, corrosion resistance, and thermal stability. Of specific interest are the effects of the primary coolant helium, which contains trace impurities of hydrogen, methane, carbon monoxide, and water vapor, on mechanical behavior. In addition, irradiation effects on the properties of absorber rod cladding and tritium permeation from the primary coolant into the process gas are important areas for investigation. The results show that, for test times of up to 20,000 h, the creep-rupture strength in air and in HTR helium lies in the same scatter band. The results of low cycle fatigue tests indicate a beneficial effect of HTR helium on the cycles of failure. Investigations of corrosion in HTR helium have shown that acceptable corrosion resistance can be achieved by strict control of the impurity content of the helium. Using the available creep-rupture data and the linear damage accumulation rule, the acceptable service lives of intermediate heat exchanger tubes were calculated for Inconel alloy 617 at 9500 C. The data that are being accumulated from the various test programs will form the basis of a design code for nuclear components operating at temperatures greater than or equal to 8000 C

  11. Advances in high temperature chemistry

    CERN Document Server

    Eyring, Leroy

    1969-01-01

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

  12. The Indian high temperature reactor programme

    International Nuclear Information System (INIS)

    Bhabha Atomic Research Centre (BARC), in India, is currently developing concepts of high temperature nuclear reactors capable of supplying process heat at a temperature around 873-1273K. These nuclear reactors are being developed with the objective of providing energy to facilitate combined production of hydrogen, electricity, and drinking water. Under the programme, currently India is developing a Compact High Temperature Reactor (CHTR) as a technology demonstrator for associated technologies. CHTR is mainly 233U-thorium fuelled, lead-bismuth cooled and beryllium oxide moderated reactor. This reactor, initially being developed to generate about 100 kW(th) power, will have a core life of around 15 years and will have several advanced passive safety features to enable its operation as compact power pack in remote areas not connected to the electrical grid. The reactor is being designed to operate at 1273K, to facilitate demonstration of technologies for high temperature process heat applications such as hydrogen production by splitting water through high efficiency thermo-chemical process. Molten lead based coolant has been selected for the reactor so as to achieve a higher level of safety. For this reactor, developmental work in the areas of fuel, structural materials, coolant technologies, and passive systems are being done in BARC. Experimental facilities are being set up to demonstrate associated technologies. In parallel, design work has been initiated for the development of a 600 MW(th) High Temperature Reactor for commercial hydrogen production by high temperature thermo-chemical water splitting processes. Technologies being developed for CHTR would be utilized for the development of this reactor. Various analytical studies have been carried out in order to compare different options as regards fuel configuration and coolants. Initial studies carried out indicate selection of pebble bed reactor configuration with either lead or molten salt-based cooling by

  13. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  14. High Temperature reactors status 1977

    International Nuclear Information System (INIS)

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

  15. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    International Nuclear Information System (INIS)

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  16. High temperature gas-cooled reactor technology

    International Nuclear Information System (INIS)

    The high temperature gas-cooled reactor (HTGR) with a direct cycle helium system has drawn attention as the next generation nuclear power plant that is closest to commercialization. Fuji Electric participated in the design, manufacture and construction of JAPCO's Tokai-1 plant, a 'Colder Hall' type reactor, which was the first commercial nuclear power plant in Japan, and JAERI's high temperature engineering test reactor (HTTR), which was the first high temperature gas-cooled reactor in Japan. Fuji Electric, a pioneer of gas-cooled reactors, worked on the design, construction and development of these reactors. This paper provides brief descriptions of the air-cooled spent fuel storage system of the HTTR, material test facilities for the HTTR, and the development of an inherently safe and highly efficient commercial HTGR power plant as examples of Fuji Electric's recent activities in the HTGR field. (author)

  17. Resonance integral calculations for high temperature reactors

    International Nuclear Information System (INIS)

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

  18. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    International Nuclear Information System (INIS)

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  19. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  20. Helium-cooled high temperature reactors

    International Nuclear Information System (INIS)

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg

  1. Advanced high temperature heat flux sensors

    Science.gov (United States)

    Atkinson, W.; Hobart, H. F.; Strange, R. R.

    1983-01-01

    To fully characterize advanced high temperature heat flux sensors, calibration and testing is required at full engine temperature. This required the development of unique high temperature heat flux test facilities. These facilities were developed, are in place, and are being used for advanced heat flux sensor development.

  2. Gas turbine high temperature reactor, GTHTR-300

    International Nuclear Information System (INIS)

    The high temperature gas reactor (HTGR) has some characters without previously set reactors such as capability of taking out heat with high temperature, high specific safety, and so on. The gas turbine high temperature reactor (GTHTR) activating such characters has some advantages such as high power generation efficiency, feasibility on simplification of safety apparatus, and so on, and that has excellent economical efficiency. Recently, this GTHTR system is positively promoted on its investigation in South Africa, U.S.A., Russia, Holland, China, France, and so on. In JAERI, on a base of the feasibility study on GTHTR carried out fiscal year 1996 to 2000 as an entrusted research by the Science and Technology Agency, a design investigation on an actual use GTHTR (GTHTR-300) with excellent safety economical efficiency and operation feature and about 300 MW in electric output by using Japanese own technology has been progressed. The GTHTR-300 is an excellent system adopted Japanese initiative also for GTHTR as well as activated some reactor related technologies accumulated on HTGR R and D in Japan at a center of HTTR (high temperature engineering test reactor). Here were described on developing target, design concept, and a route to actual use of GTHTR. (G.K.)

  3. Advances in high temperature chemistry 1

    CERN Document Server

    Eyring, Leroy

    2013-01-01

    Advances in High Temperature Chemistry, Volume 1 describes the complexities and special and changing characteristics of high temperature chemistry. After providing a brief definition of high temperature chemistry, this nine-chapter book goes on describing the experiments and calculations of diatomic transition metal molecules, as well as the advances in applied wave mechanics that may contribute to an understanding of the bonding, structure, and spectra of the molecules of high temperature interest. The next chapter provides a summary of gaseous ternary compounds of the alkali metals used in

  4. A High Temperature-Tolerant and Radiation-Resistant In-Core Neutron Sensor for Advanced Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Lei [The Ohio State Univ., Columbus, OH (United States); Miller, Don [The Ohio State Univ., Columbus, OH (United States)

    2015-01-23

    The objectives of this project are to develop a small and reliable gallium nitride (GaN) neutron sensor that is capable of withstanding high neutron fluence and high temperature, isolating gamma background, and operating in a wide dynamic range. The first objective will be the understanding of the fundamental materials properties and electronic response of a GaN semiconductor materials and device in an environment of high temperature and intense neutron field. To achieve such goal, an in-situ study of electronic properties of GaN device such as I-V, leakage current, and charge collection efficiency (CCE) in high temperature using an external neutron beam will be designed and implemented. We will also perform in-core irradiation of GaN up to the highest yet fast neutron fluence and an off-line performance evaluation.

  5. A High Temperature-Tolerant and Radiation-Resistant In-Core Neutron Sensor for Advanced Reactors. Final report

    International Nuclear Information System (INIS)

    The objectives of this project are to develop a small and reliable gallium nitride (GaN) neutron sensor that is capable of withstanding high neutron fluence and high temperature, isolating gamma background, and operating in a wide dynamic range. The first objective will be the understanding of the fundamental materials properties and electronic response of a GaN semiconductor materials and device in an environment of high temperature and intense neutron field. To achieve such goal, an in-situ study of electronic properties of GaN device such as I-V, leakage current, and charge collection efficiency (CCE) in high temperature using an external neutron beam will be designed and implemented. We will also perform in-core irradiation of GaN up to the highest yet fast neutron fluence and an off-line performance evaluation.

  6. Challenges in the development of high temperature reactors

    International Nuclear Information System (INIS)

    Highlights: • Challenges for advance reactor concepts (such as VHTR and AHTR) are discussed. • Both the VHTR and AHTR design offer promising performance characteristics and potential for process heat industrial applications. • Licensing issues needs to be addressed by increasing the technical maturity level by building and operating prototype. - Abstract: Advanced reactor designs offer potentially significant improvements over currently operating light water reactors including improved fuel utilization, increased efficiency, higher temperature operation (enabling a new suite of non-electric industrial process heat applications), and increased safety. As with most technologies, these potential performance improvements come with a variety of challenges to bringing advanced designs to the marketplace. There are technical challenges in material selection and thermal hydraulic and power conversion design that arise particularly for higher temperature, long life operation (possibly >60 years). The process of licensing a new reactor design is also daunting, requiring significant data collection for model verification and validation to provide confidence in safety margins associated with operating a new reactor design under normal and off-normal conditions. This paper focuses on the key technical challenges associated with two proposed advanced reactor concepts: the helium gas cooled Very High Temperature Reactor (VHTR) and the molten salt cooled Advanced High Temperature Reactor (AHTR)

  7. Fusion reactors-high temperature electrolysis (HTE)

    International Nuclear Information System (INIS)

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 18000C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 14000C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 106 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

  8. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

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

  9. Thermal hydraulic studies of high temperature reactors

    International Nuclear Information System (INIS)

    The development of High Temperature Nuclear Reactors capable of supplying process heat at a temperature around 1273 K, is in Progress at BARC. These nuclear reactors are being developed with the objective of providing energy to facilitate combined production of hydrogen, electricity, and drinking water. The reject and waste heat in the overall energy scheme are utilised for electricity generation and desalination, respectively. Presently, technology development for a small power (100 kWth) Compact High Temperature Reactor (CHTR) capable of supplying high temperature process heat at 1273 K is being carried out. In addition conceptual details of a 10 MWth reactor supplying heat at 1273 K for commercial hydrogen production, are also being worked out. 3D CFD analysis of the CHTR reactor core has been carried out to estimate the core heat removal capability by natural circulation during normal operating conditions. PHOENICS, a generalized CFD code is used for the analysis. The full-scale core, including fuel tube, coolant channel, plenums, down comer, heat sink, moderator and reflector has been modeled and analysed in PHOENICS. Steady state analysis is carried out to find flow distribution in the coolant circuit and temperature distribution in the whole core. Analyses have also been carried out to simulate various operational transients and accidental conditions of the reactor. This paper deals with the detailed CFD analysis. The details on the selection of the appropriate turbulence model, turbulent Prandtl number and mesh distribution for the CFD analysis are described in the paper. The results of the steady state and transient analyses are also presented in the paper. Paper shows one of the results of 3D CFD analysis for CHTR core. This paper also deals with the core thermal hydraulic analysis of the conceptual design of the 10MWth High Temperature Pebble Bed Reactor. Preliminary thermal hydraulic analysis is carried out with FLiBe as the primary coolants. The

  10. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    The report comprises three parts entitled: Dragon reactor experiment, research and development and advanced applications, and administration. As for the chapter 'Dragon-reactor experiment', the irradiation program, experimental support and post-irradiation operation are successively discussed. Chapter entitled 'Research and development and advanced applications' is dealing with fuel and graphite (kernel, coating, consolidation studies, and HTR's graphite irradiation), primary circuit materials, the power reactor physics and kinetics, and the HTR technology; the financial situation and staff of the Dragon Project at termination work are especially emphasized in third chapter

  11. A high temperature reactor for ship propulsion

    International Nuclear Information System (INIS)

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

  12. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

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

  13. The Development of an INL Capability for High Temperature Flow, Heat Transfer, and Thermal Energy Storage with Applications in Advanced Small Modular Reactors, High Temperature Heat Exchangers, Hybrid Energy Systems, and Dynamic Grid Energy Storage C

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Zhang, Xiaoqin [The Ohio State Univ., Columbus, OH (United States); Kim, Inhun [The Ohio State Univ., Columbus, OH (United States); O' Brien, James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The overall goal of this project is to support Idaho National Laboratory in developing a new advanced high temperature multi fluid multi loop test facility that is aimed at investigating fluid flow and heat transfer, material corrosion, heat exchanger characteristics and instrumentation performance, among others, for nuclear applications. Specifically, preliminary research has been performed at The Ohio State University in the following areas: 1. A review of fluoride molten salts’ characteristics in thermal, corrosive, and compatibility performances. A recommendation for a salt selection is provided. Material candidates for both molten salt and helium flow loop have been identified. 2. A conceptual facility design that satisfies the multi loop (two coolant loops [i.e., fluoride molten salts and helium]) multi purpose (two operation modes [i.e., forced and natural circulation]) requirements. Schematic models are presented. The thermal hydraulic performances in a preliminary printed circuit heat exchanger (PCHE) design have been estimated. 3. An introduction of computational methods and models for pipe heat loss analysis and cases studies. Recommendations on insulation material selection have been provided. 4. An analysis of pipe pressure rating and sizing. Preliminary recommendations on pipe size selection have been provided. 5. A review of molten fluoride salt preparation and chemistry control. An introduction to the experience from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory has been provided. 6. A review of some instruments and components to be used in the facility. Flowmeters and Grayloc connectors have been included. This report primarily presents the conclusions drawn from the extensive review of literatures in material selections and the facility design progress at the current stage. It provides some useful guidelines in insulation material and pipe size selection, as well as an introductory review of facility process and components.

  14. The Development of an INL Capability for High Temperature Flow, Heat Transfer, and Thermal Energy Storage with Applications in Advanced Small Modular Reactors, High Temperature Heat Exchangers, Hybrid Energy Systems, and Dynamic Grid Energy Storage C

    International Nuclear Information System (INIS)

    The overall goal of this project is to support Idaho National Laboratory in developing a new advanced high temperature multi fluid multi loop test facility that is aimed at investigating fluid flow and heat transfer, material corrosion, heat exchanger characteristics and instrumentation performance, among others, for nuclear applications. Specifically, preliminary research has been performed at The Ohio State University in the following areas: 1. A review of fluoride molten salts' characteristics in thermal, corrosive, and compatibility performances. A recommendation for a salt selection is provided. Material candidates for both molten salt and helium flow loop have been identified. 2. A conceptual facility design that satisfies the multi loop (two coolant loops [i.e., fluoride molten salts and helium]) multi purpose (two operation modes [i.e., forced and natural circulation]) requirements. Schematic models are presented. The thermal hydraulic performances in a preliminary printed circuit heat exchanger (PCHE) design have been estimated. 3. An introduction of computational methods and models for pipe heat loss analysis and cases studies. Recommendations on insulation material selection have been provided. 4. An analysis of pipe pressure rating and sizing. Preliminary recommendations on pipe size selection have been provided. 5. A review of molten fluoride salt preparation and chemistry control. An introduction to the experience from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory has been provided. 6. A review of some instruments and components to be used in the facility. Flowmeters and Grayloc connectors have been included. This report primarily presents the conclusions drawn from the extensive review of literatures in material selections and the facility design progress at the current stage. It provides some useful guidelines in insulation material and pipe size selection, as well as an introductory review of facility process and

  15. Reactor core design of Gas Turbine High Temperature Reactor 300

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been designing Japan's original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h. This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan

  16. Reliability Analysis of High Temperature Reactor Fuels

    International Nuclear Information System (INIS)

    This paper presents the results of reliability analysis of the TRISO -coated fuel particles for the High Temperature Test Reactor (HTTR), Japan. The reliability of fuel particle was evaluated based on the failure probability of each coating layer, and only the failure due to internal gas pressure and shrinkage of pyrolytic carbon (PyC) layer was considered The analysis results show that, no significant failure occurs up to about 45 MWd/kgU for the first core fuel particle and up to about 75 MWd/kgU for the reload core fuel particle. The fuel particle is predicted to fail completely at about 50 MWd/kgU for the first core fuel particle and at about 85 MWd/kgU for the reload core fuel particle. This results show that the TRISO -coated fuel particle for the HTTR to have high reliability. No failure occurs up to the maximum burnup design level, i.e. 33 MWd/kgU for the first core fuel particle and 60 MWd/kgU for the reload core fuel particle. The analysis results show also that the fuel particle reliability (coating layers) depends on the irradiation temperature. The failure occurs at lower burnup if the irradiation temperature increases. (author)

  17. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    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

  18. Fort St. Vrain high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The construction, testing, and preliminary operating experience of the Fort St. Vrain Nuclear Generating Station are described. This station utilizes the advanced high-temperature gas-cooled reactor (HTGR) concept and is the first nuclear reactor system in the United States to use a prestressed concrete reactor vessel (PCRV). Helium is used as the primary coolant, and a nitrogen system provides refrigeration for the low temperature equipment of the helium purification system and for the moisture monitors in the primary coolant system. Design, construction and testing to date at this station have made a significant contribution to the HTGR concept for central station electric generating plants to supply the increasing demands for electrical energy. (U.S.)

  19. Thorium fueled high temperature gas cooled reactors. An assessment

    International Nuclear Information System (INIS)

    The use of thorium as a fertile fuel for the High Temperature Gas Cooled Reactor (HTR) instead of uranium has been reviewed. It has been concluded that the use of thorium might be beneficial to reduce the actinide waste production. To achieve a real advancement, the uranium of the spent fuel has to be recycled and the requested make-up fissile material for the fresh fuel has to be used in the form of highly-enriched uranium. A self-sustaining fuel cycle may be possible in the HTR of large core size, but this could reduce the inherent safety features of the design. (orig.)

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

    International Nuclear Information System (INIS)

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

  1. Hydrogen production from high temperature electrolysis and fusion reactor

    International Nuclear Information System (INIS)

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

  2. Gas conduction in a high temperature reactor

    International Nuclear Information System (INIS)

    The hot gas is distributed and mixed by polygonal blocks in the reactor floor. Radial and an annular channel are used for this purpose. This annular channel also carried circular ducts distributed evenly over the circumference, which lead to a heat exchanger. Temperature differences across the crossection of the reactor floor are evened out by multiple deflection, combination and renewed splitting of the gas flows. (DG)

  3. High Temperature Wear of Advanced Ceramics

    Science.gov (United States)

    DellaCorte, C.

    2005-01-01

    It was initially hypothesized that advanced ceramics would exhibit favorable high te- friction and wear properties because of their high hot hardness and low achievable surface roughness welding observed in metals does not occur in ceramics. More recent tribological studies of many nitride, carbide, oxide and composite ceramics, however, have revealed that ceramics often exhibit high friction and wear in non-lubricated, high temperature sliding contacts. A summary is given to measure friction and wear factor coefficients for a variety of ceramics from self mated ceramic pin-on-disk tests at temperatures from 25 to up to 1200 C. Observed steady state friction coefficients range from about 0.5 to 1.0 or above. Wear factor coefficients are also very high and range from about to 10(exp -5) to 10(exp -2) cubic millimeters per N-m. By comparison, oil lubricated steel sliding results in friction coefficients of 0.1 or less and wear factors less than 10(exp -9) cubic millimeters per N-m.

  4. High temperature reactor development in the Netherlands

    International Nuclear Information System (INIS)

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

  5. Seismic stability of VGM type high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The main principles of the design provision of high temperature gas cooled VGM reactors seismic stability and the results of calculations, performed by linear-spectral method are presented. (author). 1 ref., 10 figs

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

    International Nuclear Information System (INIS)

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

  7. High temperature reactors and their use in the FRG

    International Nuclear Information System (INIS)

    Various aspects of the strategy of building high temperature reactors in the FRG are discussed. The development of these reactors has a long tradition in the FRG and great sums of money are being invested in the research programme. In 1988 the AVR-15 experimental reactor is expected to be shut down in which the helium output temperature had been maintained at 950 degC for a long period of time. The THTR-300 demonstration power plant which is expected to be available at that time represents a link to further application of high temperature reactors in the FRG. A detailed description is presented of projects of further high temperature reactors with a wide range of power output. The BBC/HRB association with Swiss participation is now specifying the project of the HTR-500 reactor with a steam cycle and the delivery of technological steam. This reactor should be followed up by the construction of a reactor with an HHT gas turbine and of an HTR-PNP reactor for coal gasification. Alternatively developed are small HTR-100 universal reactors. Prospective projects also include the 80 MW modular system by KWU following up on the AVR-15 reactor. (Z.M.)

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

    International Nuclear Information System (INIS)

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

  9. Design and development of high temperature heat pipes and thermosiphons for passive heat removal from compact high temperature reactor

    International Nuclear Information System (INIS)

    Compact High Temperature Reactor (CHTR) is 100 kWth, lead-bismuth eutectic (LBE) cooled reactor having several advanced passive safety features to enable its operation as compact power pack. It will also facilitate demonstration of technologies for high temperature process heat applications. In CHTR heat is transferred from primary to secondary side by means of high temperature heat pipes. Heat pipes are also employed to remove heat under postulated accident scenarios. Thus, reliable operation of heat pipes is essential for the safe working of the reactor. In this respect, computer codes have been developed for design and simulation of high temperature heat pipes. This includes design codes using empirical correlations as well as simplified FEM models for system level analysis. To verify the operation of these heat pipes under various steady state and transient conditions full CFD analysis is essential. This has been done by using a commercial CFD code by incorporating user defined functions (UDFs) which address the saturated nature of the vapour phase and the vapour wick interface conditions. A three dimensional transient numerical model has been developed to predict the vapor core, wall temperatures, vapor pressure, and vapor velocity in the screen mesh wick of sodium heat pipe. This thesis will give an outline of all the developed models and compared the predicted results against the experimental data. (author)

  10. High temperature gas-cooled reactors - perspective of thermal reactor concept with high thermal efficiency

    International Nuclear Information System (INIS)

    The present HTR development is based worldwide on the extensive experience gained in the construction and operation of gas-cooled reactors of the Magnox type and on the successful operation of the experimental high temperature reactors Dragon, Peach Bottom and AVR. The advanced CO2-cooled reactors, as well as the HTR prototype power plants for St. Vrain and THTR, are all suffering considerable delays in construction and commissioning. The commercial HTR plants have not yet achieved the decisive breakthrough onto the market. Increasing interest is being shown in advanced HTR systems, i.e., HTR with gas turbine, HTR process heat reactors and gas-cooled fast breeders. The key problem in the coming years will be the closing of the fuel cycle. Development work in this connection has already started. (orig.)

  11. Multiphysics methods development for high temperature gas reactor analysis

    Science.gov (United States)

    Seker, Volkan

    Multiphysics computational methods were developed to perform design and safety analysis of the next generation Pebble Bed High Temperature Gas Cooled Reactors. A suite of code modules was developed to solve the coupled thermal-hydraulics and neutronics field equations. The thermal-hydraulics module is based on the three dimensional solution of the mass, momentum and energy equations in cylindrical coordinates within the framework of the porous media method. The neutronics module is a part of the PARCS (Purdue Advanced Reactor Core Simulator) code and provides a fine mesh finite difference solution of the neutron diffusion equation in three dimensional cylindrical coordinates. Coupling of the two modules was performed by mapping the solution variables from one module to the other. Mapping is performed automatically in the code system by the use of a common material mesh in both modules. The standalone validation of the thermal-hydraulics module was performed with several cases of the SANA experiment and the standalone thermal-hydraulics exercise of the PBMR-400 benchmark problem. The standalone neutronics module was validated by performing the relevant exercises of the PBMR-268 and PBMR-400 benchmark problems. Additionally, the validation of the coupled code system was performed by analyzing several steady state and transient cases of the OECD/NEA PBMR-400 benchmark problem.

  12. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

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

    International Nuclear Information System (INIS)

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

  15. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  16. High temperature reactor - source of heat in new technologies

    International Nuclear Information System (INIS)

    Possible uses of nuclear energy in various industrial processes are considered. Status of development of high temperature gas-cooled reactors is given. Emphasis is given on processes of converting coal for more adequate uses of low quality lignite reserves in our country. (author)

  17. Decommissioning of the thorium high temperature reactor (THTR 300)

    International Nuclear Information System (INIS)

    The prototype Thorium-High-Temperature-Reactor (THTR 300) was decommissioned using the option of safe enclosure. Decision was made in 1989 and safe enclosure was reached in February 1997, followed by up to thirty years of operation of the safe enclosed plant. (author)

  18. Spectral emissivity of candidate alloys for very high temperature reactors in high temperature air environment

    International Nuclear Information System (INIS)

    Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values

  19. Spectral emissivity of candidate alloys for very high temperature reactors in high temperature air environment

    Energy Technology Data Exchange (ETDEWEB)

    Cao, G., E-mail: gcao@wisc.edu; Weber, S.J.; Martin, S.O.; Sridharan, K.; Anderson, M.H.; Allen, T.R.

    2013-10-15

    Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values.

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

    International Nuclear Information System (INIS)

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

  1. Heat exchanger performance in main cooling system on high temperature test operation at high temperature gas-cooled reactor 'HTTR'

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Research Institute achieved the reactor outlet coolant temperature of 950degC for the first time in the world at Apr.19, 2004. To remove generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value. (author)

  2. Analysis Of A High Temperature Gas-Cooled Reactor Powered High Temperature Electrolysis Hydrogen Plant

    International Nuclear Information System (INIS)

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322 C and 750 C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  3. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs

  4. Thermal hydraulics of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: · High temperature gas reactor fuels behavior · High temperature materials qualification · Design methods development and validation · Hydrogen production technologies · Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs. (author)

  5. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  6. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

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

  7. Hybrid high temperature gas-cooled reactor, thermonuclear fusion

    International Nuclear Information System (INIS)

    The project of a multi-purpose high temperature gas-cooled reactor started in 1969. The Atomic Energy Commission, Japan, approved in 1980 the budget for the design study of the experimental reactor. The conceptual design is in progress. The manufacturing of coated fuel pellets and the test method have been developed. The study of graphite structure is carried out. Corrosion and creep tests are made to obtain the knowledge concerning the metals in high temperature helium gas. The engineering study of various machines and structures operating at high temperature is performed. International cooperative works are considered. The experimental reactor will be critical in 1987. A critical plasma test facility, JT-60, has been constructed at the Japan Atomic Energy Research Institute. As the theoretical work on plasma confinement, the evaluation of the critical beta value of JT-60 was made. By high temperature neutral beam injection, the slowing down and heating processes of high energy particles are studied. The development of a non-circular cross-section tokamak is in progress. The construction of JT-60 will be completed in 1984. Study concerning superconducting magnets is considered. Japan is one of the members of INTOR project. (Kato, T.)

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

    International Nuclear Information System (INIS)

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

  9. High-temperature reactor developments in the Netherlands

    International Nuclear Information System (INIS)

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

  10. Present status of research and development of high temperature gas-cooled reactors, 1989

    International Nuclear Information System (INIS)

    The development of high temperature gas-cooled reactors has very important significance for the energy policy of Japan to diversify energy sources and stably ensure energy. The development has been advanced since 1969 by Japan Atomic Energy Research Institute, but in the 'Long term plan of atomic energy development and utilization' of June, 1987, as high temperature engineering test and research, it was decided to construct a High Temperature Engineering Test Reactor (HTTR). According to this policy, JAERI carried out the detailed design and safety analysis of the HTTR with 30 MWt output and reactor exit coolant temperature of 950degC. The application for the permission to construct the HTTR in Oarai Research Establishment was made in February, 1989, and it is expected to begin the construction in 1990. The period of the construction requires about 5 years, and the criticality is scheduled in 1995. In JAERI, the high temperature engineering test and research including the utilization of the HTTR and the heightening of temperature in high temperature gas-cooled reactors are promoted. The trend of development of high temperature gas-cooled reactors in foreign countries and the research activities in JAERI are reported. (K.I.)

  11. Design study on gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been conducting the design study of an original design concept of gas turbine high temperature reactor, the GTHTR300 (Gas Turbine High Temperature Reactor 300). The GTHTR300 is a greatly simplified HTGR-GT plant that leads to substantially reduced technical and cost requirements for earlier technology deployment. Also, it is expected to be an efficient and economically competitive reactor in 2010s due to newly proposed design features such as core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design. In addition, a preliminary cost evaluation proved that the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  12. High temperature fast reactor for hydrogen production in Brazil

    International Nuclear Information System (INIS)

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

  13. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

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

    International Nuclear Information System (INIS)

    matrix. The external radius of a TRISO particle has been set to 410μm, the radius of the fuel kernel to 150μm, in case of plutonium fueled core, and 215μm in case of uranium fueled core. The core was modelled in stochastic, three-dimensional code MCNP, version 4c3, in the finest detail. First, an under moderated core setup was found for both types of fuel by modifying the fuel to moderator ratio; then, the void and the thermal coefficients of reactivity were investigated. Few single - component molten salts were involved in the study of the void effect, in order to estimate worth of these components; NaF, BeF2, LiF, ZrF4. As a reference multi-component salt, Li2Be4F, referred to as FLiBe, was investigated. Results: It can be seen that removing BeF2 from the core brings a negative reactivity contribution, while other three components, NaF, LiF and ZrF4 would in a mixture contribute to the reactivity positively. Voiding FLiBe, which is a mixture of 66% of LiF and 34% BeF2, is equivalent to a negative reactivity insertion. Both the moderator and the fuel temperature coefficients of reactivity are large and negative for both plutonium and uranium fueled core. In the operational temperature interval (1200 K for graphite and 1500 K for fuel), the total temperature feedback is - 7.82 pcm/K for the plutonium fueled core and -2.47 pcm/K for the uranium fueled core. This results show, that the LS-VHTR core has a potential to meet the basic safety requirements as both uranium, and spent LWR fuel burner. References: [1] D. T. Ingersoll, L. J. Ott, J. P. Renier, S. J. Ball, W. R. Corwin, C. W. Forsberg, D. F. Williams, D. F. Wilson, L. Reid, G. D. Del Cul, P. F. Peterson, H. Zhao, P. S. Pickard, E. J. Parma. Status of Preconceptual Design of the Advanced High-Temperature Reactor. (ORNL, The United States of America, Tennessee 2004). [2] A. Talamo, W. Gudowski, F. Venneri, Annals of Nuclear Energy, 31, 173-196 (2004), The burnup capabilities of the Deep Burn Modular Helium Reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-11-01

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

  16. Economics of High-Temperature Nuclear Reactors for Industrial Cogeneration

    OpenAIRE

    Hampe, Jona; Madlener, Reinhard

    2012-01-01

    The EU Emissions Trading Scheme challenges the cost-competitiveness of energy-intensive industries in Europe, and induces them to search for low-carbon alternatives for their process heat requirements, such as cogeneration or the employment of nuclear power plants. The high-temperature nuclear reactor (HTR) is a technology option that combines these two aspects. In this paper, the economic potential of using HTRs for cogeneration of industrial process heat and electricity is studied. We show ...

  17. Hydrogen production using high temperature nuclear reactors : A feasibility study

    OpenAIRE

    Sivertsson, Viktor

    2010-01-01

    The use of hydrogen is predicted to increase substantially in the future, both as chemical feedstock and also as energy carrier for transportation. The annual world production of hydrogen amounts to some 50 million tonnes and the majority is produced using fossil fuels like natural gas, coal and naphtha. High temperature nuclear reactors (HTRs) represent a novel way to produce hydrogen at large scale with high efficiency and less carbon footprint. The aim of this master thesis has been to eva...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

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

  2. High Temperature Materials Characterization and Advanced Materials Development

    International Nuclear Information System (INIS)

    The project has been carried out for 2 years in stage III in order to achieve the final goals of performance verification of the developed materials, after successful development of the advanced high temperature material technologies for 3 years in Stage II. The mechanical and thermal properties of the advanced materials, which were developed during Stage II, were evaluated at high temperatures, and the modification of the advanced materials were performed. Moreover, a database management system was established using user-friendly knowledge-base scheme to complete the integrated-information material database in KAERI material division

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-02-01

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

  4. Status of ongoing research and results: hydrogen production project for the very high temperature reactor system

    International Nuclear Information System (INIS)

    High temperature processes for large-scale production of hydrogen are being investigated as potential uses of process heat from the Very High-Temperature Reactor (VHTR) system. Working groups of technical experts are being organized to focus cooperative efforts on specific topics. Areas of cooperation include: developing and optimizing the thermo-chemical water splitting processes of the sulphur family, giving priority to the sulphur-iodine (S-I) cycle; advancing the high-temperature electrolysis process; evaluating alternative thermo-chemical hydrogen-generation processes (including processes amenable to operation with other Generation IV reactor systems); and defining and validating technologies for coupling reactors to process plants. Progress in these areas will be described in this paper

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

    International Nuclear Information System (INIS)

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 7600C (14000F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  6. Hybrid simulation of high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    A hybrid simulator was made to calculate the dynamics of high temperature gas cooled reactor(VHTR). The continuous space-discrete time (CSDT) method is applied to solve the partial differential equations of the heat transfer in the hybrid computation. By this method the error of the heat balance is decreased to less than one percent in the steady state. Though the mini computer is used for this simulator, it operates about five times faster than real time. The dynamics of VHTR are characterized by the large heat capacity of the reactor core and the long time constant. The values of these parameters are reported as the results of this calculation. The control system of the reactivity and the coolant flow rate is required to operate the reactor. The nonlinearity of VHTR which occurs in the change of flow rate are also understood quantitatively by this simulator. (author)

  7. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Qualls, A L [ORNL

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

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

  10. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  14. Supercell Depletion Studies for Prismatic High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    J. Ortensi

    2012-10-01

    The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

  15. A high temperature reactor accommodated in a cylindrical pressure vessel

    International Nuclear Information System (INIS)

    This concerns the mixing of hot gases of a high temperature reactor with spherical fuel elements which is placed in a cylindrical pressure vessel. The cooling gas is blown in from above and after passing through the pebble bed it is collected on the bottom where the graphite blocks form a kind of pillard hall. According to this invention, in front of each channel through which the hot gas is blown off there is placed a specially formed replacement body for deviation purpose, and thus the hot gases are thoroughly mixed. Like this the gas flow gets to the components of the primary coolant circuit with a regular temperature. (GL)

  16. Increasing the power of the high temperature reactor module

    International Nuclear Information System (INIS)

    To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoretically below 1600 C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%. ((orig.))

  17. Mastery of the plutonium in high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F.; Raepsaet, X. [CEA Saclay, Dept. Modelisation des Systemes et Structures, DM2S, 91 - Gif-sur-Yvette (France); Lecomte, M. [FRAMATOME, 92 - Paris-La-Defence (France)

    2001-07-01

    Whatever the future of the civil nuclear programme in France may be, the plutonium reprocessing and recycling option has been chosen 14 years ago and the control of the plutonium inventory appears today as a major R and D issue. Many studies in progress at Cea attempt to improve plutonium recycling in PWR by increasing the amount of plutonium fed in the core, using inert matrix, new design. Moreover, in spite of their good performances and safe behaviour, innovative reactor concepts considered at the present time must also demonstrate their capacity to use at best the plutonium matter that represents at the same time a great energetic potential and strong radio-toxic source in spent fuel. In this context and with regard to the renewed interest in the High Temperature Gas-cooled Reactor (HTGR) concept, the problem of the mastery of the plutonium stock with the help of the HTGR has been undertaken at Cea in collaboration with Framatome. (author)

  18. Nuclear district-heating system with high-temperature reactor

    International Nuclear Information System (INIS)

    A nuclear district-heating system is made up of the following subsystems: nuclear heat-and-power plant, long-distance transfer line, interconnected district heating system, ultimate distribution system to the consumer, and fossil-fired peak-load and standby heating plants. The resons for selecting a district-heating feed temperature of 180degC and the annual performance curve for the interconnected system are presented. The reactor building for the high-temperature reactor is shown and a non-integrated prestressed-concrete vessel is proposed. Burst protection is provided for the coaxial duct and for the steam generator pressure vessel. The costs involved in generating district-heating supplies in the nuclear heat-and-power station are given and they are charged to thermal energy generation in the nuclear heat-and-power station. (M.S.)

  19. Seismic study on high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    The resistance against earthquakes of a high temperature gas-cooled reactor (HTGR) core with block-type fuel is not yet fully ascertained. Seismic studies must be made if such a reactor plant is to be installed in the areas with frequent earthquakes. The experimental and analytical studies for the seismic response of the HTGR core were carried out. First, the fundamental behavior, such as the softening characteristic of a single stacked column (which is piled up with blocks) and the hardening characteristic with the block impact were clarified from the seismic experiments. Second, the displacement and the impact characteristics of the two-dimensional vertical core and the two-dimensional horizontal core were studied from the seismic experiments. Finally, analytical methods and computer programs for the seismic response of HTGR cores were developed. (author) 57 refs

  20. High-temperature nuclear reactors - economy, ecology, technology

    International Nuclear Information System (INIS)

    Various advantages of the high-temperature reactor are shown. Even though it represents an improvement upon the conventional light-water reactor, this process is only slowly being introduced on the market. This situation is likely to change in the future because of the necessity to save and to use more efficiently energy reserves, nuclear as well as fossil. The sharp price increase of fossil fuel together with the dependence of the industrialised countries upon this type of energy will favor the use of other energy sources. The HTR with its 232-Th fuel cycle and its possibility to be used as a multipurpose heat source is likely to play an important role despite public resistance to nuclear energy in general and some recent drawbacks in its development. (Auth.)

  1. Mastery of the plutonium in high temperature reactor

    International Nuclear Information System (INIS)

    Whatever the future of the civil nuclear programme in France may be, the plutonium reprocessing and recycling option has been chosen 14 years ago and the control of the plutonium inventory appears today as a major R and D issue. Many studies in progress at Cea attempt to improve plutonium recycling in PWR by increasing the amount of plutonium fed in the core, using inert matrix, new design. Moreover, in spite of their good performances and safe behaviour, innovative reactor concepts considered at the present time must also demonstrate their capacity to use at best the plutonium matter that represents at the same time a great energetic potential and strong radio-toxic source in spent fuel. In this context and with regard to the renewed interest in the High Temperature Gas-cooled Reactor (HTGR) concept, the problem of the mastery of the plutonium stock with the help of the HTGR has been undertaken at Cea in collaboration with Framatome. (author)

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

    International Nuclear Information System (INIS)

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

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

    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.

  4. Radioactivity evaluation code system for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    A code system for the evaluation of the behavior of radioactive fission products (FP) in high temperature gas-cooled reactors (HTGR) is described. The first half of this report is devoted to the description of the conceivable behavior of FPs in the experimental very high temperature gas-cooled reactor being designed at JAERI. The transfer of FPs from the fuel to the primary coolant is considered in three steps; the release of FPs from the coated fuel particles; the diffusion of FPs within graphite sleeves; and the transfer of FPs from the sleeve surface to the coolant. As for the FP behavior within the primary coolant system, the deposition of FPs on various walls of the system is considered. As for the secondary and the thermal utilization systems, the transfer of tritium is specially considered. The calculation model for the transfer and deposition of fission products within the whole plant system is presented by a chart. The second half of this report describes the evaluation code system. The physical and mathematical models treated in each component code are presented and discussed. (Aoki, K.)

  5. Design of High Temperature Reactor Vessel Using ANSYS Software

    International Nuclear Information System (INIS)

    Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600oC and the outer temperature of 500 and 600oC. Result of calculation gave the maximum stress for outer temperature of 600oC was 288 N/ mm2 and strain of 0.000187. For outer temperature of 500oC the maximum stress was 576 N/ mm2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm2 can be used for vessel material with outer wall temperature of 600oC

  6. The modular pebble bed high temperature gas reactor

    International Nuclear Information System (INIS)

    Modular High Temperature Reactor power plants are characterized by the fact that standardized reactor units - modules -, each with a thermal power rating of 200-250 MW, can be interconnected to yield power plants in a broad power range. Provided that modular power plants are competitive, there is a variety of applications, e.g.: principal initial applications in the generation of electricity for a wide range of utility grid and plant sizes; co-generation of process steam and electricity, or district heat and electricity, for industrial or municipal consumers; and, in the long term, direct use of nuclear heat for process purposes e.g. gasifying coal, reforming methane etc. An essential condition for reasonably low capital costs is a simple design, taking into account the inherent safety features of small HTR's, e.g. the elimination of separate, redundant cooling systems for decay heat removal. Moreover, the safety concept must be simple, in order to minimize the engineering effort for the nuclear licensing procedure. Further, key reactor safety features should be convincingly demonstrated by full-scale test at an affordable cost, to provide a basis for standardized licensing of replicated reactors (the License-By-Test approach). In addition, the systems and structures within the nuclear envelope must be isolated such that the non-nuclear portion of the plant can be constructed as conventional power plant systems and structures. The Modular HTGR is designed to meet these conditions for safe, economical nuclear power

  7. Process heat cogeneration using a high temperature reactor

    International Nuclear Information System (INIS)

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

  8. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  9. Summary of HTGR [high-temperature gas-cooled reactor] benchmark data from the high temperature lattice test reactor

    International Nuclear Information System (INIS)

    The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000 degree C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of the measurement of k∞ using the unpoisoned technique (developed by R.E. Heineman of PNL) was subjected to extensive peer review within PNL and the General Atomic Company. A number of experiments were conducted at PNL in the Physical Constants Testing Reactor (PCTR) using both the unpoisoned technique and the well-established null reactivity technique that substantiated the equivalence of the measurements by direct comparison. Records of all data from fuel fabrication, the reactor experiments, and the analytical results were compiled and maintained to meet applicable quality assurance standards in place at PNL. Sensitivity of comparisons between measured and calculated k∞(T) data for various HTGR lattices to changes in neutron cross section data, graphite scattering kernel models, and fuel block loading variations, were analyzed by PNL for the Electric Power Research Institute. As a part of this effort, the fuel rod composition in the dilute 233UO2-ThO2 HTGR central cell (HTLTR Lattice number-sign 3) was sampled and analyzed by mass spectrometry. Values of k∞ calculated for that lattice were about 5% higher than those measured. Trace quantities of sodium chloride were found in the fuel rod that were equivalent to 22 atom parts-per-million of natural boron

  10. Development of Ni-Cr-W superalloys for high temperature components in high temperature gas-cooled reactors, 2

    International Nuclear Information System (INIS)

    Research and development have been carried out on the new superalloys as a component material for process heating high temperature gas-cooled reactors with coolant outlet temperatures of around 1000degC. The program aims at developing new superalloys which are well balanced in high temperature strength, corrosion resistance, producibility, weldability, etc. This report, the second series of development of Ni-Cr-W superalloys, is describing the results of the qualification tests performed by the Subcommittee on Advanced Superalloys from FY 1987 to FY 1990 for the evaluation of the fourth, the fifth and the sixth experimental alloys after the interim report, i.e., the first series. Based on the obtained results, it has been concluded that Ni - 18 to 19 mass % Cr - 20 to 22 mass % W - 0.03 mass % C - 0.08 mass % Ti - 0.02 to 0.05 mass % Zr - 0.002 to 0.007 mass % Y - 0.0035 to 0.006 mass % B is the optimum chemical composition. (author)

  11. Advanced High-Temperature Engine Materials Technology Progresses

    Science.gov (United States)

    1997-01-01

    The objective of the Advanced High Temperature Engine Materials Technology Program (HITEMP) at the NASA Lewis Research Center is to generate technology for advanced materials and structural analysis that will increase fuel economy, improve reliability, extend life, and reduce operating costs for 21st century civil propulsion systems. The primary focus is on fan and compressor materials (polymer-matrix composites - PMC's), compressor and turbine materials (superalloys, and metal-matrix and intermetallic-matrix composites - MMC's and IMC's), and turbine materials (ceramic-matrix composites - CMC's). These advanced materials are being developed in-house by Lewis researchers and on grants and contracts.

  12. Corrosion Issues of High Temperature Reactor Structural Metallic Materials

    International Nuclear Information System (INIS)

    Cooling helium of high temperature reactors (HTRs) is expected to contain a low level of impurities: oxidizing gases and carbon-bearing species. Reference structural materials for pipes and heat exchangers are chromia former nickel base alloys, typically alloys 617 and 230. And as is generally the case in any high temperature process, their long term corrosion resistance relies on the growth of a surface chromium oxide that can act as a barrier against corrosive species. This implies that the HTR environment must allow for oxidation of these alloys to occur, while it remains not too oxidizing against in-core graphite. First, studies on the surface reactivity under various impure helium containing low partial pressures of H2, H2O, CO, and CH4 show that alloys 617 and 230 oxidize in many atmosphere at intermediate temperatures (up to 890-970 degrees C, depending on the exact gas composition). However when heated above a critical temperature, the surface oxide becomes unstable. It was demonstrated that at the scale/alloy interface, the surface oxide interacts with the carbon from the material. These investigations have established an environmental area that promotes oxidation. When exposed in oxidizing HTR helium, alloys 617 and 230 actually develop a sustainable surface scale over thousands of hours. On the other hand, if the scale is destabilized by reaction with the carbon, the oxide is not protective anymore, and the alloy surface interacts with gaseous impurities. In the case of CH4-containing atmospheres, this causes rapid carburization in the form of precipitation of coarse carbides on the surface and in the bulk. Carburization was shown to induce an extensive embrittlement of the alloys. In CH4-free helium mixtures, alloys decarburize with a global loss of carbon and dissolution of the pre-existing carbides. As carbides take part in the alloy strengthening at high temperature, it is expected that decarburization impacts the creep properties. Carburization and

  13. Siting evaluation of High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    It is necessary to evaluate hypothetical accident to judge the appropriateness of reactor siting condition. Hypothetical accident is postulated assuming the occurrence of an accident which is unlikely to occur from a technical standpoint. The safety characteristics and/or advantages of the HTGRs are (1) slow response to core heatup events and (2) high temperature that fuel can sustain before the initiation of fission product release. A double-ended rupture of coaxial double pipe of the primary cooling system was selected as the hypothetical accident of the HTTR. Since the HTTR is a HTGR, the core temperature changes slowly and no instantaneous failure of coated fuel particles occur. Therefore, time-dependent release model was newly introduced to calculate the release amount of core contained fission products during the accident. From the result based on the analytical model developed here, appropriateness of siting condition of the HTTR was confirmed

  14. High temperature reactor and application to nuclear process heat

    International Nuclear Information System (INIS)

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

  15. Development of very high temperature reactor design technology

    International Nuclear Information System (INIS)

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

  16. Development of Very High Temperature Reactor Design Technology

    International Nuclear Information System (INIS)

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

  17. Ceramic matrix composites -- Advanced high-temperature structural materials

    International Nuclear Information System (INIS)

    This symposium on Ceramic Matrix Composites: Advanced High-Temperature Structural Materials was held at the 1994 MRS Fall Meeting in Boston, Massachusetts on November 28--December 2. The symposium was sponsored by the Department of Energy's Office of Industrial Technology's Continuous Fiber Ceramic Composites Program, the Air Force Office of Scientific Research, and NASA Lewis Research Center. Among the competing materials for advanced, high-temperature applications, ceramic matrix composites are leading candidates. The symposium was organized such that papers concerning constituents--fibers and matrices--were presented first, followed by composite processing, modeling of mechanical behavior, and thermomechanical testing. More stable reinforcements are necessary to enhance the performance and life of fiber-reinforced ceramic composites, and to ensure final acceptance of these materials for high-temperature applications. Encouraging results in the areas of polymer-derived SiC fibers and single crystal oxide filaments were given, suggesting composites with improved thermomechanical properties and stability will be realized in the near future. The significance of the fiber-matrix interface in the design and performance of these materials is evident. Numerous mechanical models to relate interface properties to composite behavior, and interpret test methods and data, were enthusiastically discussed. One issue of great concern for any advanced material for use in extreme environments is stability. This theme arose frequently throughout the symposium and was the topic of focus on the final day. Fifty nine papers have been processed separately for inclusion on the data base

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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/GWea is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GWea, respectively. (author)

  20. Safeguards concept for the THTR-300 Thorium High Temperature Reactor

    International Nuclear Information System (INIS)

    The nuclear power plant in Hamm, Federal Republic of Germany is a plant with a high temperature reactor. The fuel elements are spheres with a diameter of 60 mm, an enrichment of 93%, a 235U content of 0.96 g and a thorium content of 10.2 g. The facility is divided into two material balance areas (MBA 1: fresh fuel and spent fuel storage; MBA 2: loading facility, reactor core and discharge facility) in order to increase the transparency of the fuel element flow, and because a direct physical inventory of the core is not possible. The inventory of the core is determined by the quantity received and the quantity withdrawn, which are to be reported to Euratom. The quantity received in the core will be verified by the inspectors using a special sampling machine. The quantity withdrawn from the core is counted by independent counters at the exit of the core. At the exit of the spent fuel storage a monitoring and logging system verifies all the drums leaving the storage. The route of the spent fuel drums is continuously observed by video systems until they are finally packed into shipping containers which are to be sealed by the operator using electronic seals. (author)

  1. High-Temperature Engineering Test Reactor door valve monitor system

    International Nuclear Information System (INIS)

    This manual describes the detector design features, performance, and operating characteristics of the High-Temperature Engineering Test Reactor (HTTR) Door Valve Monitor System spent-fuel monitor. The HTTR Door Valve Monitor System (HDVM) is installed in the HTTR door valve to provide unattended monitoring data for the transfer of spent fuel through the door valve on the top of the reactor. The system includes a pair of detectors to provide direction of travel and redundancy. The fission product gamma rays are measured using ion chambers (ICs) and the curium neutrons are measured using shielded 3He detectors. There are two ICs and one 3He tube inside each detector package. Gamma-ray and neutron detector (GRAND) electronics supply power to the ICs and 3He tubes, and the data are collected in the GRAND and the Field Works computer. The system is designed to operate unattended with data pickup by the inspectors on a 90-day period. This manual gives the performance and calibration procedures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-01

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

  3. Analysis of a loss of forced cooling test using the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950degC (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and

  4. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  5. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  6. Construction of VHTRC (Very High Temperature Reactor Critical Assembly)

    International Nuclear Information System (INIS)

    This report describes the design, the safety analyses and the results of main pre-operation tests of VHTRC (Very High Temperature Reactor Critical Assembly) which has been constructed by the modification of the critical assembly, SHE (Semi-Homogeneous Experiment). The VHTRC is aimed at a 1/2 scale mock up of the experimental VHTR in the second detailed design stage. The three main features of VHTRC are that 1) the core is made of graphite blocks, and 2) the core is loaded with the coated particle fuel compacts using low enriched uranium, and that 3) the core including the graphite reflector can be heated up to 210 deg C using the electric heaters. The assembly is designed to keep the aseismatic strength of 0.3 G acceleration in both horizontal and vertical directions even at the core temperature, 210 deg C. The integrities of every components are investigated by the safety analyses and are proved by the pre-operation tests. On 13, May 1985, a basic core reached critical point for the first time. The experimental analysis showed that the critical mass calculated with the SRAC code system was only 3 % lower than the experimental value. This fact confirms that the VHTRC has been constructed very precisely within the design criteria and that the SRAC code system can give accurate results for the basic core configuration. (author)

  7. Seismic monitoring system in 10 MW high temperature reactor

    International Nuclear Information System (INIS)

    The seismic monitoring system is set up to preserve the 10MW High Temperature Reactor running normally or can be safely shut-down under the seismic status. The system is deigned for unattended station. This system is composed of high sensibility accelerometer, high reliability industry control unit (ICU) and ripe software for seismic analysing. It also uses new techniques in seismic data processing. Nine time history acceleration signals in three positions are provided to this system. The peak accelerate value will be computed out after the system confirmed the seismic condition. The master-control room will get alarm signal with sound and light as the peak accelerate value above the threshold value. The system will also calculate computing response spectrum and design response spectrum. After the comparison with this two value and theory response spectrum, a final seismic report will be printed. With performance test on the MTS vibrating table of Tongji University, It has shown that I type seismic design bases can be carried out with such system. The performance of this system is proved to be better than the assembly unit of many other seismic monitoring instruments

  8. High temperature material characterization and advanced materials development

    International Nuclear Information System (INIS)

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

  9. High temperature material characterization and advanced materials development

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-15

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

  10. Helium turbine power generation in high temperature gas reactor

    International Nuclear Information System (INIS)

    This paper presents studies on the helium turbine power generator and important components in the indirect cycle of high temperature helium cooled reactor with multi-purpose use of exhaust thermal energy from the turbine. The features of this paper are, firstly the reliable estimation of adiabatic efficiencies of turbine and compressor, secondly the introduction of heat transfer enhancement by use of the surface radiative heat flux from the thin metal plates installed in the hot helium and between the heat transfer coil rows of IHX and RHX, thirdly the use of turbine exhaust heat to produce fresh water from seawater for domestic, agricultural and marine fields, forthly a proposal of plutonium oxide fuel without a slight possibility of diversion of plutonium for nuclear weapon production and finally the investigation of GT-HTGR of large output such as 500 MWe. The study of performance of GT-HTGR reduces the result that for the reactor of 450 MWt the optimum thermal efficiency is about 43% when the turbine expansion ratio is 3.9 for the turbine efficiency of 0.92 and compressor efficiency of 0.88 and the helium temperature at the compressor inlet is 45degC. The produced amount of fresh water is about 8640 ton/day. It is made clear that about 90% of the reactor thermal output is totally used for the electric power generation in the turbine and for the multi-puposed utilization of the heat from the turbine exhaust gas and compressed helium cooling seawater. The GT-Large HTGR is realized by the separation of the pressure and temperature boundaries of the pressure vessel, the increase of burning density of the fuel by 1.4 times, the extention of the nuclear core diameter and length by 1.2 times, respectively, and the enhancement of the heat flux along the nuclear fuel compact surface by 1.5 times by providing riblets with the peak in the flow direction. (J.P.N.)

  11. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

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

    International Nuclear Information System (INIS)

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

  13. Design and present status of high-temperature engineering test reactor

    International Nuclear Information System (INIS)

    The Japan Atomic Energy commission (JAEC) decided to construct the high-Temperature engineering Test Reactor (HTTR) in 1987 for establishing and upgrading the basic technologies for advanced HTGRs and serving an irradiation test facility for research in high temperature technologies. The HTTR is a graphite-moderated and helium-gas-cooled test reactor with thermal output of 30MW and inlet and maximum outlet coolant temperature of 395 C and 950 C respectively. Construction started in March 1991 at Oarai site of the Japan Atomic Energy Research Institute (JAERI), with its first criticality at the end of 1997 to be followed after a series of functional tests of half a year. Fabrication of reactor pressure vessel, an intermediate heat exchanger (IHX), gas circulators and other main cooling components has been finished in their factories and installed to the site in 1994. At present, the construction of HTTR reactor building and installation of containment vessel, main and auxiliary cooling systems, etc. are almost completed. This paper describes design of the HTTR reactor cooling system, control system and present status of the HTTR construction

  14. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    International Nuclear Information System (INIS)

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it

  15. Numerical Simulation of Accident Scenario in High Temperature Gas Cooled (Pebble Bed) Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peter, Geoffrey J. [Oregon Institute of Technology - Portland Center, Portland (United States)

    2012-03-15

    The accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors is considered in this paper. The next generation of Advanced High Temperature Reactors (AHTR), are considered for nuclear power production, and for high-temperature hydrogen production using nuclear reactors to reduce the carbon footprint. Blockages can cause LOCA variations in flow and heat transfer that may lead to hot spots within the bed that could compromise reactor safety. Therefore, it is important to know the void fraction distribution and the interstitial velocity field in the packed bed. The blockage for this numerical study simulated a region with significantly lower void than that in the rest of the bed. Finite difference technique solved the simplified continuity, momentum, and energy equations. Any meaningful outcome of the solution depended largely upon the validity of the boundary conditions. Among them, the inlet and outlet velocity profiles required special attention. Thus, a close approximation to these profiles obtained from an experimental set-up established the boundary conditions. This paper presents the development of the elliptic-partial equation for a bed of a bed of pebbles, and the solution procedure. The paper also discusses velocity and temperature profiles obtained from both numerical and experimental set-up, with and without effect of blockage. Based on the studies it is evident that knowledge of LOCA velocity and temperature distribution within the fuel element in a Pebble Bed Nuclear Reactor or AHTR is essential for reactor safety.

  16. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    International Nuclear Information System (INIS)

    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to ∝1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  17. 3-D space time kinetics of compact high temperature reactor with fuel temperature feedback

    International Nuclear Information System (INIS)

    The Compact High Temperature Reactor (CHTR) is being developed as technology demonstrator for Indian High Temperature Reactor programme. Physics design of conceptual core of (Th-233U) fuelled CHTR is in advance stage and various core configurations have been proposed. Reactor core operation at high temperature necessitates sophisticated safety and anticipated transients analyses including postulated LORA, LOCA, and power set-back transients in CHTR. Recently, efficient IQS module in ARCH with adiabatic fuel temperature feedback capability has been developed. For accounting fuel and coolant temperature feedbacks in the simulation of 3D space time transients in CHTR, module for 1D (radial) heat conduction based module for heat transfer from fuel to coolant has been incorporated in 3D space-time analysis code ARCH. The AER benchmarking results of ARCH-IQS code with Doppler feedback and results of anticipated transient without scram (ATWS) of (Th-233U) fuelled CHTR with the present capability in ARCH-IQS code have been presented in this paper. (author)

  18. Modular High-Temperature Gas-Cooled Reactor (MHTGR) status

    International Nuclear Information System (INIS)

    The MHTGR is an advanced reactor concept being developed under a cooperative program involving the US government, the nuclear industry, and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant, and a graphite moderator. However, the specific size and configuration is selected to utilize the inherent safety characteristics associated with these standard features to develop passive safety systems that provide a significantly higher margin of safety and investment protection than current generation reactors. The design meets the Protective Action Guideline (PAG) limits at the 425 m site boundary, hence precluding the need for evacuation or sheltering of the public during any licensing basis event. The safe behavior is not dependent upon operator action and is insensitive to operator error. The conceptual design is presently being reviewed by the Nuclear Regulatory Commission (NRC). A safety evaluation report and licensability statement are scheduled to be issued by the NRC in January 1988. Status of the design with respect to applications, performance/operation, siting flexibility, investment protection, safety and economics are presented in this paper

  19. Modular high-temperature gas-cooled reactor (MHTGR) status

    International Nuclear Information System (INIS)

    The MHTGR is an advanced reactor concept being developed under a cooperative program involving the U.S. Government, the nuclear industry ,and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant, and a graphite moderator. However, the specific size and configuration is selected to utilize the inherent safety characteristics associated with these standard features to develop passive safety systems that provide a significantly higher margin of safety and investment protection than current generation reactors. The design meets the Protection Action Guideline (PAG) limits at the 425 m site boundary, hence precluding the need for evacuation of sheltering of the public during any licensing basis event. The safe behavior is not dependent upon operator action and is insensitive to operator error. The conceptual design is presently being reviewed by the Nuclear Regulatory Commission (NRC). A safety evaluation report and licensability statement are scheduled to be issued by the NRC in January 1988. Status of the design with respect to applications, performance/operation, siting flexibility, investment protection, safety and economics are presented in this paper

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

    International Nuclear Information System (INIS)

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

  1. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  2. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    International Nuclear Information System (INIS)

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540 C and the helium coolant was delivered at 7 MPa at 625-925 C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the high temperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

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

    International Nuclear Information System (INIS)

    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

  4. High temperature electrochemistry related to light water reactor corrosion

    International Nuclear Information System (INIS)

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

  5. Advancement of High Temperature Black Liquor Gasification Technology

    Energy Technology Data Exchange (ETDEWEB)

    Craig Brown; Ingvar Landalv; Ragnar Stare; Jerry Yuan; Nikolai DeMartini; Nasser Ashgriz

    2008-03-31

    Weyerhaeuser operates the world's only commercial high-temperature black liquor gasifier at its pulp mill in New Bern, NC. The unit was started-up in December 1996 and currently processes about 15% of the mill's black liquor. Weyerhaeuser, Chemrec AB (the gasifier technology developer), and the U.S. Department of Energy recognized that the long-term, continuous operation of the New Bern gasifier offered a unique opportunity to advance the state of high temperature black liquor gasification toward the commercial-scale pressurized O2-blown gasification technology needed as a foundation for the Forest Products Bio-Refinery of the future. Weyerhaeuser along with its subcontracting partners submitted a proposal in response to the 2004 joint USDOE and USDA solicitation - 'Biomass Research and Development Initiative'. The Weyerhaeuser project 'Advancement of High Temperature Black Liquor Gasification' was awarded USDOE Cooperative Agreement DE-FC26-04NT42259 in November 2004. The overall goal of the DOE sponsored project was to utilize the Chemrec{trademark} black liquor gasification facility at New Bern as a test bed for advancing the development status of molten phase black liquor gasification. In particular, project tasks were directed at improvements to process performance and reliability. The effort featured the development and validation of advanced CFD modeling tools and the application of these tools to direct burner technology modifications. The project also focused on gaining a fundamental understanding and developing practical solutions to address condensate and green liquor scaling issues, and process integration issues related to gasifier dregs and product gas scrubbing. The Project was conducted in two phases with a review point between the phases. Weyerhaeuser pulled together a team of collaborators to undertake these tasks. Chemrec AB, the technology supplier, was intimately involved in most tasks, and focused primarily on the

  6. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  7. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

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

    International Nuclear Information System (INIS)

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

  9. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Experimental investigations on the migrational behaviour of silver in coated particle fuel for high-temperature reactors

    International Nuclear Information System (INIS)

    The migrational behaviour of silver in the coated particle fuel proposed for High-Temperature Reactors, is investigated experimentally. Data are described in the framework of the diffusion model. The diffussion coefficients are derived from the experimental data by a nonlinear least squares fit procedure. The experimental procedures and the theoretical calculations to analyse the data are described extensively. Arrhenius lines are presented for U(Th)-02, PyC and Sic. The silver release in advanced High-Temperature Reactors is prognosticated based on the measured data. (orig./HP)

  12. Reactor core of a gas-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    In order to increase the outlet temperature of the coolant (helium) leaving the reactor core of a gas-cooled high-temperature nuclear reactor and thereby to improve its thermal efficiency there is proposed to design the geometry of the fuel elements or the fuel element units and/or their main dimensions non-uniformly. Those fuel elements whose geometry causes a larger pressure drop of the coolant gas are to be arranged towards the outlet side of the hot coolant. (GL) 891 GL/GL 892 MKO

  13. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  14. Analysis of the control of the high temperature gas-cooled reactor nuclear power plants

    International Nuclear Information System (INIS)

    Modular High Temperature Gas-Cooled Reactor (MHTGR) is characterized by inherent safety and higher electrical efficiency, so it can effectively improve the safety and economics of the nuclear power plants. Based upon these advantages, the High Temperature Gas-Cooled Reactor-Pebble Bed Module (HTR-PM) is under design and will be constructed in China to demonstrate the safety and economics of MHTGR. The automatic control system is important and necessary to the safe, economical, and efficient operation of the MHTGR. This paper investigates the control characteristics of the HTGR nuclear power plants, and analyzes the control technique and existing control strategies of HTGR plants. Advanced control technology which applies modern and intelligent control theory in industrial process provides an opportunity to improve the control performance of the MHTGR plant. Based upon the advanced control technology, the paper proposes a preliminary design concept of hierarchical coordinated control system to the control system design of the HTR-PM which employs the Distributed Control System (DCS) principle. (author)

  15. In-reactor optical dosimetry in high-temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    The applicability of fused silica core optical fibres to in-reactor dosimetry was demonstrated at elevated temperatures and a special irradiation rig was developed for realizing high-temperature optical dosimetry in a high-temperature test reactor (HTTR) at the Oarai Research Establishment of JAERI (Japan Atomic Energy Research Institute). The paper will describe the present status of preparation for the high-temperature dosimetry in HTTR, utilising radiation-resistant optical fibres and radioluminescent materials. Temperature measurement with a high-speed response is the main target for the present optical dosimetry, which could be applied for monitoring transient behaviours of the HTTR. This could be realised by measuring the intensity of thermoluminescence and black body radiation in the infrared region. For monitoring reactor powers, optical measurements in the visible region are essential. At present, the measurement of the intensity of Cerenkov radiation is the most promising area of study. Other possibilities with radioluminescent materials having luminescent peaks in the visible region are under consideration. One of the candidates will be silica, which has a robust radioluminescent peak at 450 nm. (author)

  16. High-temperature membrane reactors : potential and problems

    NARCIS (Netherlands)

    Saracco, G.; Neomagus, H.W.J.P.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours, describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. Since then, progress has

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

    Energy Technology Data Exchange (ETDEWEB)

    Shiozawa, Shusaku [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

  19. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-15

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

  20. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    International Nuclear Information System (INIS)

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

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

    2011-03-01

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

  3. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

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

    2010-02-23

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

  4. Environmental aspects of MHTGR [Modular High-Temperature Gas-Cooled Reactor] operation

    International Nuclear Information System (INIS)

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept being developed under a cooperative program involving the US Government, the utilities and the nuclear industry. This plant design utilizes basic High Temperature Gas-Cooled Reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The MHTGR design approach leading to exceptional safety performance also leads to plant operation which is characterized by extremely low radiological emissions even for very low probability accidents. Coated fuel particles retain radionuclides within the fuel, thus minimizing material contamination and personnel exposure. The objective of this paper is to characterize radioactive effluents expected from the normal operation of an MHTGR. In addition, other nonradioactive effluents associated with a power generating facility are discussed. Nuclear power plants produce radioactive effluents during normal operation in gaseous, liquid and solid forms. Principal sources of radioactive waste within the MHTGR are identified. The manner in which it is planned to treat these wastes is described. Like other reactors, the MHTGR produces nonradioactive effluents associated with heat generation and chemical usage. However, due to the MHTGR's higher efficiency, water usage requirements and chemical discharges for the MHTGR are minimized relative to other types of nuclear power plants. Based upon prior operating HTGR experience and analysis, effluents are quantified in terms of radioactivity levels and/or emission volume. Results, quantified within the paper, demonstrate that effluents from the MHTGR are well below regulatory limits and that the MHTGR has a minimal impact upon the public and the environment. 14 refs., 2 figs., 4 tabs

  5. Concept on inherent safety in high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    A new safety concept in a high-temperature gas-cooled reactor (HTGR) was proposed to provide the most advanced nuclear reactor that exerts no harmful consequences on the people and the environment even if multiple failures in all safety systems occur. The proposed safety concept is that the consequence of the accidents is mitigated by the confinement of fission products employing not multiple physical barriers as in light water reactors, but only the cladding of fuel (i.e., the coating layers of the coated fuel particle). The progression of the events that lead to the loss or degradation of the confinement function of the coating layers (i.e., core heat up, oxidation of the coating layers, and explosion of carbon monoxide) is suppressed by only physical phenomena (i.e., the Doppler effect, thermal radiation and natural convection, formation of a protective oxide layer for coating layers of fuel, oxidation of carbon monoxide) that emerge deterministically as a cause of the events. The feasibility studies for severe events and related information revealed that the HTGR design based on this safety concept is technically feasible. This concept indicates the direction in which nuclear reactor research should be headed in terms of safety after the accident at the Fukushima Daiichi Nuclear Power Plant. (author)

  6. Alcohol synthesis in a high-temperature slurry reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  7. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  8. Interactions of lithium with high temperature reactor materials

    International Nuclear Information System (INIS)

    Liquid lithium is used as coolant for reactors operated at temperatures greater than 1000degC. Furthermore, refractory metal (e.g. Nb., Ta or Mo) alloys have been selected for this type of application. This paper describes the corrosion and mass transfer of some selected alloys in liquid lithium. Interstitial element (e.g oxygen, nitrogen and carbon) transport and its relationship to the non-metallic impurities in the coolant are described in detail. (author)

  9. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T.

    1992-01-01

    In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL's contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

  10. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T.

    1992-01-01

    In this project we intend to study a novel process concept, i.e, the use of ceramic membranes reactors in upgrading of coal derived liquids. Membrane reactors have been used in a number of catalytic reaction processes in order to overcome the limitations on conversion imposed by thermodynamic equilibrium. They have, furthermore, the inherent capability for combining reaction and separation in a single step. Thus they offer promise for improving and optimizing yield, selectivity and performance of processes involving complex liquids, as those typically found in coal liquid upgrading. Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sol-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

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

    International Nuclear Information System (INIS)

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

  12. The high temperature reactor in the future fuel market

    International Nuclear Information System (INIS)

    The commercial GA 1160 MW(e) station reactor is designed to operate on the thorium cycle with recycling of the bred 233U. The low-enriched uranium fuel cycle can be accommodated in the existing design without major alterations to the reactor and the shutdown system. The reference thorium reactor operating in the 233U recycle mode is 10 to 20% cheaper than the low-enriched reactor; however, the thorium cycle depends on the supply of 93% enriched uranium and the availability of reprocessing and refabrication facilities to utilise its bred fissile material. A step towards realising the closing of the thorium cycle lies in simplifying the head-end reprocessing technology by abandoning the segregation concept of feed and breed coated particles in the reference cycle. A one-coated-particle scheme in which all discharged uranium isotopes are recycled in mixed oxide particles is feasible and suffers only a very minor economic penalty. Towards the end of the century there will presumably be pronounced increases in the cost of natural resources. In the case of nuclear energy, resource considerations are reflected in the price of uranium, which is expected to have reached 50 $/lb U3O8 in the early 1990's and 100 $/lb U3O8 around 2010. The fuel cycle advantage of the thorium system amounts to some 30% and is capable of absorbing substantial expenses in bringing about the closing of the out-of-pile cycle. A most attractive aspect for the HTR fuel cycle flexibility is for the utility to start operating the reactor on the low enriched uranium cycle and at a later date switch over to the thorium cycle as this becomes economically more and more attractive. As a result of detailed investigations this option is demonstrated for the GA 1160 MW design. The transition phase will cause no larger problems than those encountered in the normal operation of either of the cycles. The economic incentive amounts to some 50 Mio $ in terms of present worth money at the time of decision making

  13. High temperature metallic materials for gas-cooled reactors

    International Nuclear Information System (INIS)

    The Specialists' Meeting was organized in conjunction with an earlier meeting on this topic held in Vienna, Austria, 1981, which provided for a comprehensive review of the status of materials development and testing at that time and for a description of test facilities. This meeting provided an opportunity (1) to review and discuss the progress made since 1981 in the development, testing and qualification of high temperature metallic materials, (2) to critically assess results achieved, and (3) to give directions for future research and development programmes. In particular, the meeting provided a form for a close interaction between component designers and materials specialists. The meeting was attended by 48 participants from France, People's Republic of China, Federal Republic of Germany, Japan, Poland, Switzerland, United Kingdom, USSR and USA presenting 22 papers. The technical part of the meeting was subdivided into four technical sessions: Components Design and Testing - Implications for Materials (4 papers); Microstructure and Environmental Compatibility (4 papers); Mechanical Properties (9 papers); New Alloys and Developments (6 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. This volume contains all papers presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  14. New fuel design for high temperature gas reactor

    International Nuclear Information System (INIS)

    Since the end of 2000, the CEA is committed in an international collaboration for the research and development of the future nuclear reactors of generation IV. To achieve the characteristics of the new concepts of reactor, different designs of fuel have been settled. This paper deals with the so-called 'honeycomb matrix' in which the fuel is spread out. This concept seems to be very promising on account of many advantages: the honeycomb structure can offer a very close confinement of the fission gas products; this kind of structure offers a large free volume up to 90% of total volume; the fabrication of the honeycomb matrix is completely broken away from the whole manufacture of the fuel and realized out of nuclear fence. This paper includes two parts: the first part describes the main steps of the manufacturing of silicon carbide honeycombs structures by extrusion process and microstructural characterizations; the second part describes a process for the filling of the matrix by a simulated material, in the perspective of the realization of uranium compounds filled honeycombs. (authors)

  15. Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

    International Nuclear Information System (INIS)

    The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation. A discussion is pesented of the history and some of the results of this effort with respect to the advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The nonlinear dynamic analysis techniques employed to analyze the HTGR core are described

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

    International Nuclear Information System (INIS)

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

  17. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  18. High temperature indentation tests on fusion reactor candidate materials

    International Nuclear Information System (INIS)

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 oC. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 oC. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods

  19. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T.

    1992-06-19

    Ceramic membranes are a new class of materials, which have shown promise in a variety of industrial applications. Their mechanical and chemical stability coupled with a wide range of operating temperatures and pressures make them suitable for environments found in coal liquid upgrading. In this project we will evaluate the performance of Sel-Gel alumina membranes in coal liquid upgrading processes under realistic temperature and pressure conditions and investigate the feasibility of using such membranes in a membrane reactor based coal liquid upgrading process. In addition, the development of novel ceramic membranes with enhanced catalytic activity for coal-liquid upgrading applications, such as carbon-coated alumina membranes, will be also investigated.

  20. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  1. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital

  2. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

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

    International Nuclear Information System (INIS)

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

  4. High temperature gas-cooled reactors - Operating on fuel recycle

    International Nuclear Information System (INIS)

    The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power-density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. Under recycle uranium assumptions, the resource utilization is particularly attractive due to the high neutronic value of the bred, and recycled, U-233 produced from the thorium irradiation. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U3O8 requirement of 3030 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U3O8 requirement by about 10%. Additional resource utilization improvements of 11 to 14% could be realized by improved fuel management techniques

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

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

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

  6. New approach to handle neutron startup sources in a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In a high temperature gas-cooled reactor, a neutron startup source (NS) cannot be handled simply. An appropriate transfer vessel, connected with a 241Am-Be source for startup core physics tests through an NS coupler and drive wire, was newly designed and installed in the high temperature engineering test reactor (HTTR). A result of tests using the HTTR revealed that the NS could be loaded simply and certainly into the reactor core through the transfer vessel below the allowable limit of effective dose equivalent for operators

  7. The high temperature reactor - an interim balance of development and operation

    International Nuclear Information System (INIS)

    The high temperature reactor is the modern version of the gas cooled reactor. The interim balance presented in this report therefore refers to all gas cooled types of reactor, i.e. Magnox, AGR, and HTR. The period covered by the balance begins 28 years ago, when the British gas cooled reactor of Calder Hall went critical for the firt time. The experience accumulated with the German experimental HTR plant, the Juelich AVR reactor, has been extremely satisfactory, both with respect to the operating behaviour and to safety. The HTR fuel elements have a high safety margin against excessive operating temperatures. Although the dominating role played by the light water reactor line has so far prevented the commercial application of high temperature reactors, developments in recent years seem to indicate new market chances for the high temperature reactor line. In this connection, special importance attaches to the prototype THTR-300, which is about to be commissioned, and to the HTR-100 and HTR-500 conceptual design drafts and the modular reactor. As the design data of the THTR-300 and the HTR-500 are partly identical, the latter plant is characterized by foreseeable preparation and construction times; in addition, the licensability of the HTR-500 has already been confirmed. A medium sized reactor like this could be the link between electricity generation and the generation of process heat and space heat. (orig.)

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

    International Nuclear Information System (INIS)

    Power plants can be divided into conventional steam plants, fueled with hard coal, lignite or liquid fuel, hydroelectric plants and nuclear plants, their chief use was or is the production of electric energy and - in certain cases only - of production of process heat, using steam or hot water for process heat in industry and district heating for residential and commercial purposes. The part played by electricity in the whole energy demand is of the order of 10% to 25% the total demand, the rest is necessary for supplying process heat below 2000C or above 2000C, up to some 15000C. The present distribution of energy demands is covered chiefly by liquid fuel, coal and lignite, water energy and increasing steps by nuclear fuel. It is well known that the erection of nuclear energy plants is a necessity for today and for the future. There is another necessity, i.e. to utilize the primary energy resources in a complex way i.e. to supply electricity as energy vector and other fuels as process heat as new energy vectors. These manmade fuels - whether in a gaseous or liquid phase - contain hydrogen, and one can believe, the world is entering a new energy civilisation in utilizing hydrogen and its compounds as second energy vector. The author has taken up the task to investigate this new problem of process, heat in the form of hydrogen and its compounds, by evaluating their present and future production, based on the utilization of natural gas, oil coal, water and the nuclear heat of helium, available in a closed circuit as primary coolant in a High - Temeprature Helium cooled reactor, which is symbolized in the paper as HTR. The paper deals in more detail with the following application of Nuclear Heat: hydrogasification, direct reduction of ore, mainly iron ores, ammonia synthesis, methanol synthesis Hydrocracking, long distance transfer of process heat (chemical heat pipe), hydrogenation of coal, Fischer - Tropsch synthesis, oxosynthesis, coal gasification, coal

  9. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  10. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

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

  14. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  15. Integrity assessment of test fuel assemblies of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Assessment of integrity has been made on the B-type fuel assemblies, which will be loaded in the High Temperature Engineering Test Reactor (HTTR) as test fuel assemblies. Specifications of coated fuel particles for the B-1 type fuel assembly have been slightly changed in the fuel kernel diameter and thickness of coating layers from those for the A-type fuel assembly, which is employed as the driver fuel. These changes have been directed toward safer side in developing this advanced fuel for use up to higher burnups at higher temperatures. The B-2 type fuel assembly uses the zirconium-carbide (ZrC) coating layer with excellent high-temperature chemical stability, instead of the silicon carbide (SiC) layer. This change has lead to demonstration of its better performance than the A-type fuel assembly in the kernel migration, corrosion by fission products including palladium, and coating failure at extremely high temperatures. The B-3 type fuel assembly adopts the (U,Th)O2 kernel - SiC TRISO coated fuel articles. The service condition (1000degC and 22,000 MWd/t) of the B-3 type fuel assembly is decided as the range within which the performance data of the fuel have been sufficiently obtained. Thus, it has been judged that the integrity of these B-type fuel assemblies will be maintained under the normal operating conditions of the HTTR. Moreover, the validity of the permissible design limit of the fuel has been confirmed, which requires that the fuel temperature shall not exceed 1,600degC at anticipated operational transients. (author)

  16. Research high-temperature reactor (to thirtieth anniversary of energetic start up)

    International Nuclear Information System (INIS)

    The historical aspects of creating the first national high-temperature gas-cooled reactor - the unique experimental reactor for testing the reactor rocket-propelled engines are stated. The complex of unordinary requirements to the reactor demanded the search and realization of multiple physical, design and technological innovations, solution of complicated scientific-technological problems both in the process of development and startup process. The data on the reactor design and characteristics, its auxiliary stand systems, as well as the results of the first startup and subsequent multiyear operation are presented

  17. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  18. Performance of advanced high-temperature fuels for nuclear propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Stark, W.A.; Butt, D.P.; Storms, E.K.; Wallace, T.C. [Los Alamos National Lab., NM (United States)

    1994-12-31

    Nuclear propulsion using hydrogen has been demonstrated to operate at nearly twice the performance level of today`s chemical rockets. However, higher temperatures lead to a variety of degradations that compromise safety and longevity. Foremost among these is the melting of the propulsion reactor fuel. The melting behaviour of the U-Zr-C and U-Nb-C systems have been evaluated.

  19. The investigation for attaining the optimal yield of oil shale by integrating high temperature reactors

    International Nuclear Information System (INIS)

    This work presents a systemanalytical investigation and shows how far a high temperature reactor can be integrated for achieving the optimal yield of kerogen from oil shale. About 1/3 of the produced components must be burnt out in order to have the required high temperature process heat. The works of IGT show that the hydrogen gasification of oil shale enables not only to reach oil shale of higher quality but also allows to achieve a higher extraction quantity. For this reason a hydro-gasification process has been calculated in this work in which not only hydrogen is used as the gasification medium but also two high temperature reactors are integrated as the source of high temperature heat. (orig.)

  20. Advanced targeted monitoring of high temperature components in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E.; Maile, K.; Jovanovic, A. [MPA Stuttgart (Germany)

    1998-12-31

    The article presents the idea of targeted monitoring of high-temperature pressurized components in fossil-fueled power plants, implemented within a modular software system and using, in addition to pressure and temperature data, also displacement and strain measurement data. The concept has been implemented as a part of a more complex company-oriented Internet/Intranet system of MPA Stuttgart (ALIAS). ALIAS enables to combine smoothly the monitoring results with those of the off-line analysis, e. g. sensitivity analyses, comparison with preceding experience (case studies), literature search, search in material databases -(experimental and standard data), nonlinear FE-analysis, etc. The concept and the system have been implemented in real plant conditions several power plants in Germany and Europe: one of these applications and its results are described more in detail in the presentation. (orig.) 9 refs.

  1. Super-critical carbon dioxide based brayton cycle for Indian High Temperature Reactors

    International Nuclear Information System (INIS)

    The most effective way to improve economic competitiveness of NPPs is to enhance its efficiency which has remained static at around 33% since the first commercial LWR came into operation. New generation reactor designs including the six Gen-IV reactor concepts aim to increase the NPPs efficiency to almost 50%. This is proposed to be achieved by high temperature designs using Brayton cycle based power conversion systems. World over, Super-critical Carbon dioxide Brayton Cycle (SCBC) for power generation is an important R and D area. High efficiency SCBC power conversion system is proposed as power conversion system for Indian Molten Salt Breeder Reactor (IMSBR) and Innovative High Temperature Reactor (IHTR). This section provides the details regarding design and development of SCBC for these reactors. (author)

  2. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  3. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  4. Development of the maintenance technologies for the future high-temperature gas cooled reactor (HTGR) using operating experiences acquired in high-temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    This paper describes the lessons learned of the maintenance technologies, which have been and will be developed by using the high-temperature engineering test reactor (HTTR), which should be expected to apply to the future high-temperature gas cooled reactors (HTGRs). For example, the periodical maintenance for the future HTGRs is planned to carry out in each two years. The duration of periodical maintenance is planned about 60 days to satisfy the availability of operation up to 90%, in which decay heat removal, refueling and maintenances of equipment and components are carried out. As the key issue is to make a practical application of the technologies to the future HTGRs, shortening the duration of the periodical maintenance is expected to be important by excluding the maintenance related to the reactor equipment from the critical path by shifting the time-based maintenance, which is defined to be carried out in certain time interval, to the condition-based maintenance, which is defined to be carried out when a parameter exceeds its criteria, by using the experience and data acquired in HTTR, which is expected to make the maintenance interval significantly longer than that of the time-based maintenance as the operation experiences are accumulated. The maintenance technologies for the reactor system, which have been developed by using HTTR, are categorized as the following. (1) Establishment of the maintenance technologies specific to the HTGRs. (2) Development of the maintenance technologies for the future HTGRs. (3) Efficient maintenance works for the general equipment. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

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

  6. Status of structural design codes for high-temperature gas-cooled reactor components

    International Nuclear Information System (INIS)

    Beginning with basic principles for the structural design of metallic HTR components, the special features of an HTR are compared with those of light-water reactor components. The main emphasis is on components for elevated and high temperature application. The materials under consideration are introduced. The typical high-temperature failure modes are described and design limits are proposed. The methods for avoiding different types of failure are provided. The HTR concept of integrity excludes spontaneous large-area failure; therefore, the understanding of high-temperature fracture mechanics and the ongoing experimental and theoretical work, which established the main topic for the structural design code, are discussed. (orig.)

  7. Economic outlook for HTR reactors in the high temperature process heat market

    International Nuclear Information System (INIS)

    The economic aspects of using nuclear power for process steam are examined and an economic analysis is made of the markets for high temperature steam. It is shown that the circumstances which favour high temperature reactors are the largest possible size in dual purpose plants, high steam/electricity ratio, for small or medium-sized reactors, relatively high fossil fuel costs, and the highest possible load factor. The most promising application in the short and mean term appears to be hydrogen or H2 + CO production for use in the chemical, petrochemical, and iron and steel industries. (U.K.)

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

    International Nuclear Information System (INIS)

    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)

  9. Development of a high temperature, high sensitivity fission counter for liquid metal reactor in-vessel flux monitoring

    International Nuclear Information System (INIS)

    Advanced liquid metal reactor concepts such as the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Inherently Safe Module (PRISM) have relatively large pressure vessels that necessitate in-vessel placement of the neutron detectors to achieve adequate count rates during source range operations. It is estimated that detector sensitivities of 5 to 10 counts/center dot/s/center dot//sup /minus/1//center dot/[neutron/(cm2/center dot/s)]/sup /minus/1/ will be required for the initial core loading. The Instrumentation and Controls Division of Oak Ridge National Laboratory has designed and fabricated a fission counter to meet this requirement which is also capable of operating in uncooled instrument thimbles at primary coolant temperatures of 500 to 600/degree/C. Components are fabricated from Inconel-600, and high temperature alumina insulators are employed. The transmission line electrode configuration is utilized to minimize capacitive loading effects

  10. Technical outline of a high temperature pool reactor with inherent passive safety features

    International Nuclear Information System (INIS)

    Many reactor designers world wide have successfully established technologies for very small reactors (less than 10 MWTH), and technologies for large power reactors (greater than 1000 MWTH), but have not developed small reactors (between 10 MWTH and 1000 MWth) which are safe, economic, and capable of meeting user technical, economic, and safety requirements. This is largely because the very small reactor technologies and the power reactor technologies are not amiable to safe and economic upsizing/downsizing. This paper postulates that new technologies, or novel combinations of existing technologies are necessary to the design of safe and economic small reactors. The paper then suggest a set of requirements that must be satisfied by a small reactor design, and defines a pool reactor that utilizes lead coolant and TRISO fuel which has the potential for meeting these requirements. This reactor, named LEADIR-PS, (an acronym for LEAD-cooled Integral Reactor, Passively Safe) incorporates the inherent safety features of the Modular High Temperature Gas Cooled Reactor (MWGR), while avoiding the cost of reactor and steam generator pressure vessels, and the safety concerns regarding pressure vessel rupture. This paper includes the description of a standard 200MW thermal reactor module based on this concept, called LEADIR-PS 200. (author)

  11. State of development and development trends of power stations with high temperature reactors

    International Nuclear Information System (INIS)

    At present, no nuclear power plant equipped with a high temperature reactor is suitable for being introduced shortly into the European or US market, promising to rival the 1,300 MW power plants, even if several plants had been built offering all the advantages of the high temperature reactor. Variants are analyzed, and the question is dealt with, whether spherical fuel elements, the non-integrated method of construction and the prestresssed cast iron pressure vessel would not, after all, offer better chances for its competitive capacity when taking into account all the requirements to be met by the reactor system. The decision concerning the choice of the reactor system is to be prepared in the course of the next months. Since such a project with its whole fuel cycle included exceeds the financial power of the individual ocmpanies, and perhaps even that one of the individual nations, close international co-operation is in the offing. (orig.)

  12. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    International Nuclear Information System (INIS)

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold (1) efficient low cost energy generation and (2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  13. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  14. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    International Nuclear Information System (INIS)

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat ã l'Energie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  15. Analytical chemistry requirements for advanced reactors

    International Nuclear Information System (INIS)

    The nuclear power industry has been developing and improving reactor technology for more than five decades. Newer advanced reactors now being built have simpler designs which reduce capital cost. The greatest departure from most designs now in operation is that many incorporate passive or inherent safety features which require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. India is developing the Advanced Heavy Water Reactor (AHWR) in its plan to utilise thorium in nuclear power program

  16. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    In this study the economic feasibility of new concepts of the High Temperature Reactor were investigated. These new concepts are characterized as inherently safe. The different concepts were used as industrial heat/power reactors and compared with a gas fired Steam and Gas turbine installation. The best economic advantages are offered by a HTR with a Thorium/Uranium cycle as compared with a gas fired steam- and gas turbine. 6 figs, 9 tabs, 21 refs

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

    OpenAIRE

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R&D. The development of HTR started in the 1950s as a graphite moderated and helium cooled reactor. This concept featured important inherent...

  18. The key device--elevator in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The basic structure, working principle and behavior of the control system of the elevator in 10 MW high temperature gas-cooled reactor (HTR-10) are researched. The five-phase hybrid stepping motor and the closed-loop control are adopted in the construction design of the elevator. About 20000 fuel elements and graphite balls were transported into the reactor core by the elevator to achieve the critical loading for HTR-10

  19. Generic Investigations on Transport Theory Modelling of High Temperature Reactors of Pebble Bed Type

    OpenAIRE

    Sureda Sureda, Antonio Jaime

    2008-01-01

    The GRS (Gesellschaft fuer Anlagen und Reaktorsicherheit = Company for Plant and Reactor Safety) maintains and further develops the code system DORT-TD/HERMIX-DIREKT, which is a complex tool for the simulation of coupled neutronics/thermal-hydraulics transients and accident scenarios of high-temperature gas cooled reactors of pebble bed type. With this tool, GRS takes part in the international benchmark activity "OECD/NEA PBMR400 Transient Benchmark”, which aims at the simulation of transient...

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

    International Nuclear Information System (INIS)

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

  1. Critical experiments and reactor physics calculations for low enriched high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    On the recommendation of the IAEA International Working Group on Gas Cooled Reactors, the IAEA established a Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low-Enriched High Temperature Gas Cooled Reactors (HTGRs) in 1990. The objective of the CRP was to provide safety-related physics data for low-enriched uranium (LEU) fueled HTGRs for use in validating reactor physics codes used by the participating countries for analyses of their designs. Experience on low-enriched uranium, graphite-moderated reactor systems from research institutes and critical facilities in participating countries were brought into the CRP and shared among participating institutes. The status of experimental data and code validation for HTGRs and the remaining needs at the initiation of this CRP were addressed in detail at the IAEA Specialists Meeting on Uncertainties in Physics Calculations for HTGR Cores held at the Paul Scherrer Institute (PSI), Villigen, Switzerland in May, 1990. The main activities of the CRP were conducted within an international project at the PROTEUS critical experiment facility at the Paul Scherrer Institute, Villigen, Switzerland. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Fuel for the experiments was provided by the KFA Research Center, Juelich, Germany. Initial criticality was achieved on July 7, 1992. These experiments were conducted over a range of experimental parameters such as carbon-to-uranium ratio, core height-to-diameter ratio, and simulated moisture concentration. To assure that the experiments being conducted are appropriate for the design of the participants, specialists from each of the countries have participated

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-09-01

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

  4. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  5. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  6. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

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

    International Nuclear Information System (INIS)

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si3N4. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

  8. Fuel handling system of 10 MW high temperature gas cooling reactor based on LabVIEW

    International Nuclear Information System (INIS)

    The field multi-channel signals has been acquired synchronously from 10 MW High temperature gas cooling reactor fuel handling system by DAQ technology. Counting software is developed based on LabVIEW. Its virtual instrument is flexible and user-friendly, and can count fuel-ball exactly. (authors)

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

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable f

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

    International Nuclear Information System (INIS)

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

  11. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  12. Design of the material performance test apparatus for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor (HTGR). To solve the problem, a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR. The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus. Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field. A typical experimental process was analyzed in the paper, which lasted 76 hours including seven stages. Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified. Test temperature in the apparatus could be elevated up to 1600℃, which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor. (authors)

  13. High-temperature gas-cooled reactors (HTGRs) and their potential for non-electric application

    International Nuclear Information System (INIS)

    This paper presents High Temperature Gas cooled Reactors (HTGR). It also enumerates the potentials for non electrical applications such as delivering hot water, generating steam, producing hydrogen and carbon monoxide via conversion of natural gas. Then the author presents the contribution of HTGRs to reduce carbon dioxide emissions. (TEC). 4 figs., 1 ref

  14. Thermal insulation of the high-temperature helium-cooled reactors

    International Nuclear Information System (INIS)

    Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper

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

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

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

  16. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Atsushi, E-mail: shimizu.atsushi35@jaea.go.jp [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Kawamoto, Taiki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Tochio, Daisuke [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Saito, Kenji; Sawahata, Hiroaki; Honma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Takada, Shoji [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Shinozaki, Masayuki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan)

    2014-05-01

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established.

  17. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    International Nuclear Information System (INIS)

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established

  18. STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh

    2011-03-01

    An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

  20. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  1. High temperature and sensitivity fission chambers: qualification of the CFUCO7 in reactor

    International Nuclear Information System (INIS)

    We present, in this paper, the whole tests performed both in laboratory and in reactor on the high temperature, wide dynamic fission chamber CFUCO7 and on its associated electronics. Except the long time tests to be realized in the PHENIX reactor, this measurement device, fission chamber and wide range electronic, can be considered as qualified to be used in a large LMFBR. We present also the new improvements on the detector design and the future programme in the reactor SUPER-PHENIX. (authors). 9 figs., 4 tabs., 2 refs., 2 appendix

  2. Reactivity-insertion-transient analysis of a fluoride salt cooled high temperature reactor

    International Nuclear Information System (INIS)

    The Fluoride salt cooled High temperature Reactor (FHR) is an innovative reactor design that uses conventional TRISO high temperature fuel with a low-pressure liquid salt coolant. The design of this reactor is currently in progress both in China and in the United States. An FHR based on ordered pebble bed core design is being planned for construction by the Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary reactivity insertion transient analysis of an FHR of SINAP's pebble core design, using RELAP5/MOD4.0 code. New models and methodologies are developed for several prototypical facilities that are based on SINAP's pebble bed concept, and different types of reactivity insertion transient are analyzed. SINAP's design is currently in progress; the ultimate goal of the transient analysis is to acquire the capability of RELAP5/MOD4.0 for performing FHR core design. (author)

  3. Thermal Assessment of Very High Temperature Reactors: Direct and Indirect Brayton Power Cycles

    International Nuclear Information System (INIS)

    Search for a sustainable energy supply system has driven nuclear engineering towards what has been termed Generation IV. One of the main objectives of these innovative nuclear designs is to reach a high thermal efficiency in their power cycles. This way a substantial fuel saving and waste reduction is achieved, which would enhance competitiveness of the nuclear kWh. This paper investigates the thermal performance of helium-cooled power cycles based on the characteristic working parameters of Very High Temperature Reactor systems both at present and at the near future. Direct (C(IC)2HTRTX) and indirect (C(IC)2(IHX)TX) cycle baselines have been modelled and both have shown an excellent thermal performance ( with thermal efficiencies near or even higher than 50%). Enhancement of associated technology would increase thermal performance (i.e., and wnet) of both cycles drastically. The analysis of the results indicate that from the thermal performance standpoint, the cycle C(IC)2HTRTX would be a better option. However, when advances in associated technologies are considered, the efficiency gap between the two baselines analyzed become smaller. In no case net power turned out to be a differential feature between the layouts. (authors)

  4. 3D AGENT methodology validation for prismatic high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The Generation IV of nuclear reactors includes as highly competitive the design of a Very High Temperature Reactor (VHTR). This type of reactors can be of a prismatic block, or pebble-bed type. An example of a prismatic block nuclear reactor is the High Temperature Test Reactor (HTTR) operated by Japan Atomic Energy Agency; the reactor reached its full power of 30 MWth for the first time in 1999. The primary coolant is helium at the pressure of ∼4 MPa, with inlet-outlet temperatures of 395°C and 850 – 950°C, respectively. The fuel is 6% enriched uranium, and the moderator is made of graphite. Using the literature available data, a comprehensive validation study is performed to benchmark and assess the AGENT (Arbitrary GEometry Neutron Transport) methodology capabilities in predicting and capturing reactor physics details affected by double heterogeneity of the fuel. Using AGENT with explicit modeling of the fuel double heterogeneity, the HTTR neutronics parameters are compared to NEWT and KENO VI, as well as to experimental data as found in literature. Detailed analysis of spatial steady-state reaction rates and flux spatial maps are provided. The AGENT methodology is based on the method of characteristics and the only one in the world as applied to reactor systems, the R-function based reactor solid modeler, in providing an accurate deterministic solution for 3D steady-state reactor physics. The R-functions modeler presents no limits to reactor geometry and materials types with their distributions. (author)

  5. Comprehensive thermal hydraulics research of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: high temperature gas reactor fuels behaviour, high temperature materials qualification, design methods development and validation, hydrogen production technologies energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  6. Evaluation of two processes of hydrogen production starting from energy generated by high temperature nuclear reactors

    International Nuclear Information System (INIS)

    In this work an evaluation to two processes of hydrogen production using energy generated starting from high temperature nuclear reactors (HTGR's) was realized. The evaluated processes are the electrolysis of high temperature and the thermo-chemistry cycle Iodine-Sulfur. The electrolysis of high temperature, contrary to the conventional electrolysis, allows reaching efficiencies of up to 60% because when increasing the temperature of the water, giving thermal energy, diminishes the electric power demand required to separate the molecule of the water. However, to obtain these efficiencies is necessary to have water vapor overheated to more than 850 grades C, temperatures that can be reached by the HTGR. On the other hand the thermo-chemistry cycle Iodine-Sulfur, developed by General Atomics in the 1970 decade, requires two thermal levels basically, the great of them to 850 grades C for decomposition of H2SO4 and another minor to 360 grades C approximately for decomposition of H I, a high temperature nuclear reactor can give the thermal energy required for the process whose products would be only hydrogen and oxygen. In this work these two processes are described, complete models are developed and analyzed thermodynamically that allow to couple each hydrogen generation process to a reactor HTGR that will be implemented later on for their dynamic simulation. The obtained results are presented in form of comparative data table of each process, and with them the obtained net efficiencies. (author)

  7. Core design and safety analyses of 600 MWt, 950 °C high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki, E-mail: nakano-m@fujielectric.co.jp [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Takada, Eiji; Tsuji, Nobumasa; Tokuhara, Kazumi; Ohashi, Kazutaka; Okamoto, Futoshi [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Tazawa, Yujiro; Tachibana, Yukio [Japan Atomic Energy Agency, Oarai, Ibaraki-pref. 311-1393 (Japan)

    2014-05-01

    The conceptual core design study of high temperature gas-cooled reactor (HTGR) is performed. The major specifications are 600 MW thermal output, 950 °C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag-110m and Cs-137 from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

  8. Material requirements for the Very High Temperature Reactor results and progress within the RAPHAEL-IP

    International Nuclear Information System (INIS)

    Full text of publication follows.The modular VHTR is one of six advanced fission systems of interest for meeting the Generation IV goals of attaining highly economic, safe, reliable, sustainable, proliferation-resistant systems. The VHTR offers significant advantages for long-term development of sustainable energy and in particular for heat applications and hydrogen generation. This system can operate with either a direct or indirect cycle and makes use of the high efficiency Brayton cycle. Work on materials investigations for the HTR within Europe recommenced with the EU 5. Framework Programme (5FP) projects HTR-M and M1 [1] and together with other SFP projects (fuel, reactor physics, components, safety,..) and the establishment of the European High Temperature Reactor Technology Network HTR-TN, served as the main European platform for the co-ordination and development of VHTR issues. The HTR-M and M1 projects addressed material requirements for the key components of the direct cycle HTR. The work especially focused on the materials development for the pressure vessel, high temperature components (including turbine), and the graphite core. Alongside this, developments were undertaken on key component issues (HTR-E) associated with the gas turbine, the recuperator and other system developments (e.g. tribology, corrosion, bearings, seals, etc.) concerned the operation and performance of the power circuit components. Within this paper the main highlights from the results of the 5FP programmes affecting material issues are reviewed and examined. For the 6. Framework Programme activities the main European research focus on VHTR is through the RAPHAEL Integrated Project (IP). The project started in 2005 and addresses a range of issues (materials, components, fuel, code qualification, etc.), which are structured in a similar way to the corresponding GIF VHTR projects. The materials issues are addressed within one of the RAPHAEL, sub-projects with a focus on outstanding

  9. Accident simulations and post irradiation examinations on spherical fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    An important aspect of the safety of high temperature reactors is the quality of the nuclear fuel and its ability to remain intact even at high temperatures and to safely contain the radioactive fission products. In combination with a suitable reactor an inherent safety against large release of fission products can be achieved. In this work experimental simulations of severe accidents were conducted on spherical fuel elements for high temperature reactors with TRISO-coated particles and fission product release was measured. The fuel elements originated from various irradiation experiments conducted at high temperatures with high burn-up. The experiments were performed using the cold finger apparatus, a test apparatus which was already used in the past in a former version at the Research Center Juelich. The new cold finger apparatus is installed since 2005 in the Hot Cells of the European Institute for Transuranium Elements. The cold finger apparatus at the Institute for Transuranium enabled incident simulations on irradiated high temperature reactor fuel elements in a helium atmosphere at ambient pressure, at temperatures up to 1800 C and for periods of several hundred hours. Here, both the release of fission gases and the release of solid fission products were measured. In addition, in the context of the present study, the mechanical behavior of the fuel particles and the transport mechanisms of the main fission products were analyzed and the expected release was computed. For a better understanding of the processes post irradiation examinations were conducted on the available fuel elements. It was finally made an assessment of the test results which were compared with results in the existing literature. A key objective of the work was the extension of the existing data base for modern HTR-fuel towards higher burn-up and higher fluences of fast neutrons, higher operating temperatures and extended accident temperatures.

  10. Design of the inner reflector of a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Graphite undergoes considerable changes in its physical and mechanical characteristics when exposed to fast neutrons and high temperatures in the High-Temperature Reactor core. The inner reflector, which is situated adjacent to the core, is designed for this type of material behaviour. Depending on the loading, the geometrical layout and the engineering design are defined by the special characteristics of graphite. The dimensional criteria are set by the integrity of the reflector blocks. In the event of extremely high neutron fluences, wear of the reflector block surface adjacent to the core is allowed. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  13. Does the age of high temperature gas-cooled reactors arrive?

    International Nuclear Information System (INIS)

    Recently, the increase of population, the expansion of resource and energy consumption, and the protection of global environment have become topics, and the expectation to atomic energy becomes large mainly in Asia. At present, the utilization of atomic energy is limited to electric power generation, and is about 17% of the total generated electric power in the world, and about 6% of the total energy consumption. The development of high temperature gas-cooled reactors has been carried out for more than 30 years, which can heighten the efficiency of power generation to 50% from 30% of LWRs, and the heat of close to 1000degC can be utilized. Also in Japan, the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output and the coolant temperature at reactor exit of 950degC is constructed in Japan Atomic Energy Research Institute, aiming at attaining the initial criticality in 1998. The features of high temperature gas-cooled reactors are low output density and large heat capacity, the negative large temperature coefficient of reactivity, and no core melting. The utilization of nuclear heat is carried out by converting to the steam up to 510degC or helium gas up to 950degC. As the present state of research and development, heat and electricity combined supply system, the methanol production by coal gasification, the hydrogen production by thermochemical process are reported. (K.I.)

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

    International Nuclear Information System (INIS)

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

  15. THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Michael G. McKellar

    2011-11-01

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

  16. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    International Nuclear Information System (INIS)

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRi-ISOtropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1,600 to 1,650 C, remaining well below the melting point of the fuel. Two key aspects associated with the safety of HTR fuel are assessed in this paper: fission product retention at temperatures up to 1,800 C is analyzed with the Cold Finger Apparatus (KueFA) while the behaviour of HTR-relevant fuel materials in an oxidizing environment is studied with the Corrosion Apparatus KORA. The KueFA is used to observe the combined effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in helium atmosphere for several hundred hours, mimicking accident temperatures up to 1,800 C and realistic temperature transients. Nongaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium (HPGe) gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. To achieve a good quantification of the release, a careful calibration of the setup is necessary and a collimator needs to be used in some cases. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Another relevant HTR accident scenario is air ingress into the reactor vessel as a consequence of a DLOFC incident. In case of

  17. DELIGHT-6(revised): one dimensional lattice burnup code for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The code, DELIGHT-6, performs the multi-group neutron spectrum calculation and provides the few-group constants for burnup calculations of a high temperature gas-cooled reactor core, whose fuel elements containing many coated fuel particles are arranged in double heterogeneity. The main revisions in the DELIGHT-6 (Revised) are as follows; (1)The option of a sphere fuel cell calculation is added for the core design of pebble bed type high temperature gas-cooled reactor. (2)The yield and decay constants of fission products for burnup calculation is revised. (3)The following auxiliary functions are added; (i) Automatic calculation of averaged atom number density in the fuel region, (ii) Estimation of local neutron flux distribution (disadvantage factor), (iii) Preparation of the data for the fine mesh core calculation. (author)

  18. Technology assessment. HTR Part 2. Application of High temperature Reactor in industrial processes

    International Nuclear Information System (INIS)

    The results of a study on the use of a high temperature reactor as a process heat reactor in high temperature (industrial) processes are presented.The study of the different temperature levels of industrial processes showed that using the heat for steam reforming of natural gas for the production of methanol and ammonia and for the use in a naphtha cracker seem to be to most promising options for the HTR. A description is given of each process and the use of the HTR heat is outlined. The consequences for process operation and operating costs using the HTR heat are discussed as well. In the case study of steam reforming energy conservation potentials in the production of methanol and ammonia are described. 20 figs., 11 tabs, 57 refs

  19. ODS Steel As A Structural Material For High Temperature Nuclear Reactors

    International Nuclear Information System (INIS)

    Oxide-dispersed-strengthened (ODS) ferritic-martensitic steels are examined as possible candidates for the structural materials to be used in the future generation of High-Temperature Gas-Cooled Nuclear Reactors, and as a replacement for alternative high-temperature materials for tubing and other structural components. ODS steels are also being considered as possible material for use in future fusion applications. Since the oxide particles serve as an interfacial pinning mechanism for moving dislocations, the creep resistance of the material is improved. However, in order to use such materials in a reactor, their behaviour under irradiation must be thoroughly examined. In this work, the effects induced by He implantation are investigated the induced swelling is measured, and the mechanical behaviour of the irradiated surface is analysed. These first tests are performed at room temperature, for which clear evidence of swelling and hardening could be observed. (author)

  20. Report of the 1st technical meeting on high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The 1st Technical Meeting on High Temperature Gas Cooled Reactors (HTGRs) was held on February 1 and 2, 1990 in the Tokai Research Establishment in order to review the results of R and D associated with the High Temperature Engineering Test Reactor (HTTR) accumulated so far in the JAERI and to investigate how to promote the further R and D on high temperature engineering and examination. From the point of view for establishing and upgrading the technology basis of HTGRs, the R and D results obtained so far and the present status of R and D were reviewed for the key items in the meeting, and the R and D items to be investigated positively and items of international cooperation to be promoted in future were discussed based on the comments and suggestions offered by the experts outside the JAERI. This report summarizes the papers which were presented on each subject of R and D in the meeting along with the comments and suggestions by the experts outside the JAERI. The results of the meeting will be reflected effectively for promoting the R and D on the high temperature engineering and examination. (author)

  1. Proceedings of the 2nd technical meeting on high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    From the point of view for establishing and upgrading the technology basis of HTGRs, the 2nd Technical Meeting on High Temperature Gas-cooled Reactors (HTGRs) was held on March 11 and 12, 1992, in Tokai Research Establishment in order to review the present status and the results of Research and Development (R and D) of HTGRs, to discuss on the items of R and D which should be promoted more positively in the future and then, to help in determining the strategy of development of high temperature engineering and examination in JAERI. At the 2nd Technical Meeting, which followed the 1st Technical Meeting held in February 1990 in Tokai Research Establishment, expectations to the High Temperature Engineering Test Reactor (HTTR), possible contributions of the HTGRs to the preservation of the global environment and the prospect of HTGRs were especially discussed, focusing on the R and D of Safety, high temperature components and process heat utilization by the experts from JAERI as well as universities, national institutes, industries and so on. This proceedings summarizes the papers presented in the oral sessions and materials exhibited in the poster session at the meeting and will be variable as key materials for promoting the R and D on HTGRs from now on. (author)

  2. Behavior of a high-temperature gas reactor with transuranic fuels

    International Nuclear Information System (INIS)

    In this work, we modeled a high-temperature gas reactor, HTGR, of prismatic block type using the SCALE 6.0 code to analyze the use of transuranic fuel in these reactors. To represent the concept, the Japanese HTTR reactor was chosen. The fuels considered used transuranic elements from UREX+ reprocessing of burned PWR fuel spiked with depleted U or Th. The calculations, performed for typical temperatures of HTR reactors, showed that, in mixtures with the same percentage of fissile material, the initial effective multiplication factor, Keff , is higher in the mixtures containing Th than that with U. Comparisons between the two types of fuel were performed using fuel pairs with the same initial Keff. During burn-up, the two mixtures show a slow and practically equal decrease in Keff. For the same level of burnup, mixtures containing Th show greater effectiveness in burning transuranics and total plutonium when compared to corresponding mixtures with depleted U. (author)

  3. Dynamic response simulation for high temperature gas-cooled reactor with indirect closed Brayton cycle

    International Nuclear Information System (INIS)

    A transient simulation program is developed in order to study dynamic characteristics of high temperature gas-cooled reactor with indirect closed Brayton cycle. After the brief introduction to such a plant, detailed mathematical models for important installations are described in the paper. By inducing step positive reactivity into the reactor, it looks like that the powers of turbo machine installations have a different growth rate accompanied with small increase of reactor power. Furthermore, this paper shows the temperature changes of reactor and heat exchangers. For the heat exchangers of the whole secondary loop, the pressure changes behave quite differently for those three sections divided by turbine, low pressure compressor and high pressure compressor. For all these equipments, the simulation program gives reasonable results and is in accordance with dynamic characteristics of their own. (authors)

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

    International Nuclear Information System (INIS)

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 9500C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

  5. Analysis of passive residual heat removal system of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The passive residual heat removal system plays an important role for the inherent safety of high temperature gas-cooled reactor (HTGR). The thermal hydraulic calculation method for the residual heat removal system of HTGR was introduced. The operating temperatures of the residual heat removal system at different residual heat powers and different environmental temperatures were calculated. The containment concrete temperature was numerically simulated. The results show that the highest concrete temperature is acceptable. (authors)

  6. The present status and future prospects of the high temperature reactor - HTR-rapporteur report

    International Nuclear Information System (INIS)

    In the context of the INFCE study, this summary report is intended to reflect the collective experience and expert opinion of the WG 8 contributors on the subject of the high-temperature, gas-cooled reactor (HTR) and its fuel cycles. The focus of interest is the role which the HTR can play in supplying future energy demands and the associated effects on resource utilization, environmental impacts, and proliferation risks

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

    International Nuclear Information System (INIS)

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

  8. Mechanical properties of structural materials for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Structural materials for high temperature gas cooled reactor should have good properties such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Mechanical properties of major structural materials, such as pressure vessel, heat exchanger, control rod, were investigated. Effect of He and irradiation on these structural materials were investigated

  9. Process for distinguishing and sorting spherical fuel elements from high temperature nuclear reactors

    International Nuclear Information System (INIS)

    Each fuel element is heated to the outlet temperature of the high temperature reactor or to a temperature of 3000 or 6000C in a furnace. Helium flows through the furnace for three minutes and takes up the fission products released. The active carbon kept at the temperature of liquid nitrogen absorbs the fission products and the Xsub(e)-133 content is then measured. (GL)

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

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

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

  11. Evaluation of reactivity coefficients for High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    This report presents the evaluation methods and evaluated results of doppler-, moderator temperature- and power coefficients for High Temperature Engineering Test Reactor (HTTR). From this study, it was made clear that the HTTR possesses inherent power-suppressing feed back characteristic due to the negative power coefficient though the moderator temperature coefficient is slightly positive due to the accumulated isotopes 135Xe and 239Pu. (author)

  12. High-temperature strain measurements on the closure studs of reactor pressure vessels

    International Nuclear Information System (INIS)

    High-temperature strain measurements have been carried out in different operating phases on two closure studs each of several reactor pressure vessels. The strains and stresses during instationary states (e.g. start-up) were of special interest. On the basis of the strains measured, it was to be proved that the calculation methods and boundary conditions on which the analytical investigation of the vessel is based are sufficiently accurate. (orig./RW)

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

    International Nuclear Information System (INIS)

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

  14. Integration Of High Temperature Gas Reactors With In Situ Oil Shale Retorting

    International Nuclear Information System (INIS)

    This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

  15. Fabrication of spherical fuel element for 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Cold quasi-isostatic molding with a silicon rubber die was used for manufacturing the spherical fuel elements of 10 MW high temperature gas-cooled reactor. 44 batches of fuel elements, about 20540 of the fuel elements, were produced. The cold properties of the graphite matrix materials satisfies the design specifications. The mean free uranium fraction in spherical fuel element from 44 batches is 4.57 x 10-5, certified products is 99%

  16. INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING

    Energy Technology Data Exchange (ETDEWEB)

    Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

    2011-05-01

    This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

  17. The AVR high-temperature reactor - operating experience, storage and final disposal of spent fuel elements

    International Nuclear Information System (INIS)

    The AVR is the first power plant with helium-cooled HTR to use spherical fuel elements. The experimental reactor was in successful operation for 21 years. In the first years of operation the main aim was the demonstration of the technical feasibility of high-temperature reactors. Special importance was attached to the testing and behavior of the fuel elements. The AVR was decommissioned in late 1988 and approve 170,000 spent fuel elements of various designs and compositions have been discharged. HTR fuel element reprocessing is not economically viable. Final disposal of the fuel elements is therefore envisaged after several years of intermediate storage. 3 refs., 1 tab

  18. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

    OpenAIRE

    Xingtuan Yang; Yanfei Sun; Huaiming Ju; Shengyao Jiang

    2014-01-01

    After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was ca...

  19. Design of a large-scale, multi-purpose high temperature gas-cooled reactor system

    International Nuclear Information System (INIS)

    The trial design of a large-scale, multi-purpose high temperature gas-cooled reactor system is described on its three aspects: nuclear reactor, nuclear heat utilization, and safety. The system is a littoral iron and steel making plant employing a multi-purpose HTGR (heat output 3,000 MW) with helium gas temperature of 1,0000C; the capacity is about 6,300,000 tons of crude steel production per year. It consists of a direct reduction furnace for ore and an electric furnace, and also an electric power generating facility. (Mori, K.)

  20. Fuel performance models for high-temperature gas-cooled reactor core design

    International Nuclear Information System (INIS)

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10-4 fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience

  1. Gas-Liquid Mass Transfer in a Slurry Bubble Column Reactor under High Temperature and

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The gas-liquid mass transfer of H2 and CO in a high temperature and high-pressure three-phase slurry bubble column reactor is studied. The gas-liquid volumetric mass transfer coefficients κLα are obtained by measuring the dissolution rate of H2 and CO. The influences of the main operation conditions, such as temperature, pressure, superficial gas velocity and solid concentration, are studied systematically. Two empirical correlations are proposed to predict κLα values for H2 and CO in liquid paraffin/solid particles slurry bubble column reactors.

  2. A project definition for demonstrating a modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    This paper describes the modular high temperature gas-cooled reactor (MHTGR) design developed within the U.S. HTGR Program and the MHTGR demonstration project recently defined by Gas-Cooled Reactor Associates (GCRA) and its utility members. A Project Definition Study was funded by GCRA and the Tennessee Valley Authority (TVA) and was cost-shared by the participating contractors. The study considered a repowering option at an existing utility site and a stand-alone, full-plant option at a remote, government site. Through the proposed demonstration project, the MHTGR would become a commercial alternative for the U.S. utility industry

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  4. In-core fuel management optimization of a Very High Temperature pebble-bed Reactor

    International Nuclear Information System (INIS)

    A new calculation procedure was developed to reduce the power peak in the core of a Very High Temperature pebble-bed Reactor. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern directly, without any intermediate loading pattern. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

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

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

    International Nuclear Information System (INIS)

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

  7. Introducing the high-temperature reactor into the market - status and strategy

    International Nuclear Information System (INIS)

    The pebble bed high-temperature reactor has been taken to the threshold of commercialization in more than 30 years of development. On the basis of the experience gained with the 15-MW AVR reactor and the THTR 300, marketable plant sizes (HTR 500, HTR Module, gas-cooled heating reactor - GHR) have been developed for the electricity and heat market, which are now available for future energy supply. The high-temperature reactor represents a reasonable supplement to the proven light-water reactors and is particularly suited for export to developing countries and industrial threshold countries in view of its technical and safety characteristics and its wide range of applications in the electricity and heat market. ABB and Siemens have decided to pursue the future HTR product development and marketing activities in a long-term strategy by the joint HTR-GmbH company. There is a worldwide interest in the HTR which is manifested by several international cooperation agreements. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 10930C (1200, 1400, 1600, 1800, and 20000F). There are a number of large industrial process heat applications that could utilize the VHTR

  12. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    similar to that used for the core of the pebble-bed advanced high-temperature reactor. This paper describes the details of the loop design, auxiliary systems used to support the facility, inductive heating system, and facility capabilities

  13. Improving high-temperature measurements in nuclear reactors with Mo/Nb thermocouples

    International Nuclear Information System (INIS)

    Many irradiation experiments performed in research reactors are used to assess the effects of nuclear radiations on material or fuel sample properties, and are therefore a crucial stage in most qualification and innovation studies regarding nuclear technologies. However, monitoring these experiments requires accurate and reliable instrumentation. Among all measurement systems implemented in irradiation devices, temperature-and more particularly high-temperature (above 1000 degrees C)-is a major parameter for future experiments related, for example, to the Generation IV International Forum (GIF) Program or the International Thermonuclear Experimental Reactor (ITER) Project. In this context, the French Commissariat a l'Energie Atomique (CEA) develops and qualifies innovative in-pile instrumentation for its irradiation experiments in current and future research reactors. Logically, a significant part of these research and development programs concerns the improvement of in-pile high-temperature measurements. This article describes the development and qualification of innovative high-temperature thermocouples specifically designed for in-pile applications. This key study has been achieved with technical contributions from the Thermocoax Company. This new kind of thermocouple is based on molybdenum and niobium thermo-elements, which remain nearly unchanged by thermal neutron flux even under harsh nuclear environments, whereas typical high-temperature thermocouples such as Type C or Type S are altered by significant drifts caused by material transmutations under the same conditions. This improvement has a significant impact on the temperature measurement capabilities for future irradiation experiments. Details of the successive stages of this development are given, including the results of prototype qualification tests and the manufacturing process. (authors)

  14. Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 deg. C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 deg. C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 deg. C, respectively, at full power operation and at the scram from the operation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

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

  16. Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2009-05-01

    A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

  17. A Safe Solution to World Energy Supply - the Very High Temperature Pebble Bed Reactor

    International Nuclear Information System (INIS)

    For the energy hungry world there is a solution which has the potential to resolve most of the present energy needs, with almost zero pollution and high thermal efficiency. The Very High Temperature Reactor (VHTR) can produce Hydrogen for automotive needs to replace the polluting gas and oil; it can produce electricity at very high efficiency with almost no pollution, and provide clean process heat for the industry and the energy needed for desalination plants to provide fresh water. In the present study it is shown that choosing the Pebble Bed concept for the VHTR is not only a very effective way to supply all the energy needs, it is also one of the safest nuclear reactor concept. Depending on the fuel cycle chosen, it is possible to reduce significantly the TRU waste normally produced in light water reactors and thus further reduce the environmental concerns of long living FP. A conceptual 600MWt High Temperature Pebble Bed reactor is proposed, and its safety characteristics are analyzed by simulating various hypothetical accidents, using the DSNP simulation system

  18. Femtosecond laser induced breakdown spectroscopy of silver within surrogate high temperature gas reactor fuel coated particles

    International Nuclear Information System (INIS)

    The detection of metallic silver on Chemical Vapour Deposited (CVD) grown silicon carbide and in Pebble Bed Modular Reactor (PBMR) supplied tri-structural isotropic (TRISO) coated particles (with 500 μm diameter zirconium oxide surrogate kernel) has been studied with femtosecond Laser Induced Breakdown Spectroscopy (femto-LIBS). The SiC layer of the TRISO coated particle is the main barrier to metallic and gaseous fission products of which 110mAg is of particular interest for direct cycle high temperature reactors. This work is a feasibility study for diagnosing and profiling silver transport through the silicon carbide layer of fuel particles for a high temperature gas reactor in out-of-reactor experimentation. The zirconium oxide is a surrogate for the enriched uranium oxide fuel. The conclusion reached in this study was that femto-LIBS can achieve good surface spatial resolution and good depth resolution for studies of silver in experimental coated particles. The LIBS technique also offers a good alternative for a remote analytical technique.

  19. Femtosecond laser induced breakdown spectroscopy of silver within surrogate high temperature gas reactor fuel coated particles

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, D.E., E-mail: troberts@csir.co.za [CSIR National Laser Centre, PO Box 395, Meiring Naude Road, Pretoria 0001 (South Africa); Plessis, A. du [CSIR National Laser Centre, PO Box 395, Meiring Naude Road, Pretoria 0001 (South Africa); Laser Research Institute, Physics Department, University of Stellenbosch, Private Bag X1, Matieland 7602 (South Africa); Steyn, J. [CSIR National Laser Centre, PO Box 395, Meiring Naude Road, Pretoria 0001 (South Africa); Botha, L.R. [CSIR National Laser Centre, PO Box 395, Meiring Naude Road, Pretoria 0001 (South Africa); Laser Research Institute, Physics Department, University of Stellenbosch, Private Bag X1, Matieland 7602 (South Africa); Strydom, C.A. [Chemical Resource Beneficiation, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Rooyen, I.J. van [PBMR, Fuel Design, 1279 Mike Crawford Avenue, Centurion, 0046 (South Africa)

    2010-11-15

    The detection of metallic silver on Chemical Vapour Deposited (CVD) grown silicon carbide and in Pebble Bed Modular Reactor (PBMR) supplied tri-structural isotropic (TRISO) coated particles (with 500 {mu}m diameter zirconium oxide surrogate kernel) has been studied with femtosecond Laser Induced Breakdown Spectroscopy (femto-LIBS). The SiC layer of the TRISO coated particle is the main barrier to metallic and gaseous fission products of which {sup 110m}Ag is of particular interest for direct cycle high temperature reactors. This work is a feasibility study for diagnosing and profiling silver transport through the silicon carbide layer of fuel particles for a high temperature gas reactor in out-of-reactor experimentation. The zirconium oxide is a surrogate for the enriched uranium oxide fuel. The conclusion reached in this study was that femto-LIBS can achieve good surface spatial resolution and good depth resolution for studies of silver in experimental coated particles. The LIBS technique also offers a good alternative for a remote analytical technique.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trond Bjornard; John Hockert

    2011-08-01

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

  2. Performance of high-temperature gas-cooled reactor as a tritium production device for fusion reactors

    International Nuclear Information System (INIS)

    Highlights: ► The performance of a gas-cooled reactor as a tritium production device was studied. ► Gas-cooled reactors with 3 GWt output can produce 5–8 kg of tritium in a year. ► Use of Li2O compound is efficient compared with Li4SiO4 or Li2TiO3 one. ► Amount of tritium produced can be increased by reducing the enrichment of 235U. - Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production device is examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the 3-dimensional entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It is shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 5–8 kg of tritium in a year.

  3. Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

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

  4. The high temperature reactor (HTR) and the new German safety concept for future nuclear power plants

    International Nuclear Information System (INIS)

    In Germany, a new law demands new safety properties for future nuclear power reactors. Nuclear reactors must be constructed in a way that even in the most unlikely accidents people in the vicinity do not have to leave their homes, i.e. the consequences must remain limited to the plant itself. This demand may be met by different technical solutions. One of these is the High Temperature Gas-cooled Reactor with spherical fuel elements as it has been developed and used for more than 20 years in Germany. The consequences of the new legal situation as well as some implications for the transfer of this nuclear power technology to developing countries are discussed. (author). 25 refs, 10 figs

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-09-01

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

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

    International Nuclear Information System (INIS)

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

  9. Helium circulator design considerations for modular high temperature gas-cooled reactor plant

    International Nuclear Information System (INIS)

    Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US

  10. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  11. Regenerated Fiber Bragg Grating sensors for high temperature monitoring in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Optical Fiber Sensors relying on high temperature resistant and spectrally multiplexed Fiber Bragg Gratings transducers are proposed for remote multi-points temperature monitoring in fourth-generation reactors such as Sodium-cooled Fast Reactors. FBGs are well-established transducers finding applications in many industrial fields. Their use is usually limited to temperature up to 400°C in continuous use. Thanks to innovative thermal engineering methods of glass at the nanometer scale, FBGs operating lifetime in high temperature environments (up to 900°C) is significantly increased, opening the way for a wide range of applications for the nuclear industry. This paper will present the thermal regeneration process used to stabilize FBGs transducers. Long-term annealing experiments up to 890°C illustrate the relevant improvements of their stability. Multiplexed regenerated FBGs have also been tested in liquid sodium up to 500°C in order to investigate their compatibility, their response time and to show potential for high temperature gradient mapping. (author)

  12. Nuclear hydrogen using high temperature electrolysis and light water reactors for peak electricity production

    International Nuclear Information System (INIS)

    In a carbon-dioxide-constrained world, the primary methods to produce electricity (nuclear, solar, wind and fossil fuels with carbon sequestration) have low operating costs and high capital costs. To minimise the cost of electricity, these plants must operate at maximum capacity; however, the electrical outputs do not match changing electricity demands with time. A system to produce intermediate and peak electricity is described that uses light water reactors (LWR) and high temperature electrolysis. At times of low electricity demand the LWR provides steam and electricity to a high temperature steam electrolysis system to produce hydrogen and oxygen that are stored. At times of high electricity demand, the reactor produces electricity for the electrical grid. Additional peak electricity is produced by combining the hydrogen and oxygen by operating the high temperature electrolysis units in reverse as fuel cells or using an oxy-hydrogen steam cycle. The storage and use of hydrogen and oxygen for intermediate and peak power production reduces the capital cost, increases the efficiency of the peak power production systems, and enables nuclear energy to be used to meet daily, weekly and seasonal changes in electrical demand. The economic viability is based on the higher electricity prices paid for peak-load electricity. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR) as moderator, thermal absorber and also as structural components of the core. 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 m3 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, studies about the

  17. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  18. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  19. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  20. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  1. Fuel particles for high temperature reactors; Combustibles a particules pour reacteurs a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pheip, M. [CEA Cadarache (DEN/CAD/DEC/SESC/LIPA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Masson, M. [CEA Valrho, Dept. Radiochimie et Procedes, 30 (France); Perrais, Ch. [CEA Cadarache (DEN/DEC/SPUA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Pelletier, M. [CEA Cadarache (DEN/DEC/SESC), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles

    2007-07-15

    The concept of fuel particles with a millimeter size was born at the end of the 1950's and is the reference concept of high or very high temperature gas-cooled reactors (HTR/VHTR). The specificity of this fuel concerns its fine divided structure, its all-ceramic composition and its micro-confining properties with respect to fission products. These 3 properties when combined together allow the access to high temperatures and to a high level of safety. This article presents: 1 - the general properties of particle fuels; 2 - the fabrication and control of fuel elements: nuclei elaboration processes, vapor deposition coating of nuclei, shaping of fuel elements, quality control of fabrication; 3 - the fuel particles behaviour under irradiation: mechanical and thermal behaviour, behaviour and diffusion of fission products, ruining mode; 4 - the reprocessing of particle fuels: stakes and options, direct storage, separation of constituents, processing of carbonous wastes; 5 - conclusion. (J.S.)

  2. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    fuels originates from goals for achieving high burnup, operating at higher temperature, and the incorporation of the minor actinides (Np, Am, Cm) into the fuels. High burn-ups will allow uninterrupted reactor operations over longer periods of time and consequently, reduction of spent fuel volumes, and eventually a significant fuel cycle reduction cost. High burn-ups are however associated with physical limitations which are primary due to the swelling of the fuel and oxidation of cladding inner surface as well as the dimensional stability of core materials such as cladding and subassembly duct due to high fast neutron dose. Higher temperature operation also challenges the performance of cladding materials and hence advanced cladding materials are needed for high temperature operation. The irradiation performance database for (U,Pu)N mixed nitride (MN) fuels is substantially smaller than that for metal carbide (MC) fuels, and these fuels can be considered to be at an early stage of development relative to oxide and metal fuels. Compared to MC fuels, MN fuels exhibit less fuel swelling, lower fission gas release, however, the problem of the production of biologically hazardous 14C in nitride fuels fabricated using natural nitrogen poses a considerable concern for the nitride spent fuel waste management. Interest remains in nitride fuels due to the combination of high thermal conductivity and high melting point. The paper also addresses the technology readiness level (TRL) concept as applied to various fuel options. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

    The annual production of hydrogen in the world is about 500 billion m3. Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

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

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been developing a hydrogen production system with a high temperature gas-cooled reactor (HTGR). While the HTGR hydrogen production system has the following advantages compared with a fossil-fired hydrogen production system; low operation cost (economical fuel cost), low CO2 emission and saving of fossil fuel by use of nuclear heat, it requires some items to be solved as follows; cost reduction of facility such as a reactor, coolant circulation system and so on, development of control and safety technologies. As for the control and safety technologies, JAERI plans demonstration test with hydrogen production system by steam reforming of methane coupling to 30 Wt HTGR, named high temperature engineering test reactor (HTTR). Prior to the demonstration test, a 1/30-scale out-of-pile test facility is in construction for safety review and detailed design of the HTTR hydrogen production system. Also, design study will start for reduction of facility cost. Moreover, basic study on hydrogen production process without CO2 emission is in progress by thermochemical water splitting. (orig.)

  6. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

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

    International Nuclear Information System (INIS)

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

  8. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  9. Modular high-temperature gas-cooled reactor core heatup accident simulations

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs

  10. Assessment of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    OpenAIRE

    Zhang, Z.; Dong, Y.; Scherer, W.

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemente...

  11. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  12. Effect of Thermal Degradation on High Temperature Ultrasonic Transducer Performance in Small Modular Reactors

    Science.gov (United States)

    Bilgunde, Prathamesh N.; Bond, Leonard J.

    Prototype ultrasonic NDT transducers for use in immersion in coolants for small modular reactors have shown low signal to noise ratio. The reasons for the limitations in performance at high temperature are under investigation, and include changes in component properties. This current work seeks to quantify the issue of thermal expansion and degradation of the piezoelectric material in a transducer using a finite element method. The computational model represents an experimental set up for an ultrasonic transducer in a pulse-echo mode immersed in a liquid sodium coolant. Effect on transmitted and received ultrasonic signal due to elevated temperature (∼200oC) has been analysed.

  13. Model analysis of influences of the high-temperature reactor on location shifting in chemical industry

    International Nuclear Information System (INIS)

    An analysis is presented of the influences of High-Temperature Reactor on probable location shifting of big chemical plants, in the future. This is done by a spatial location model, that includes an investigation on 116 industrial locations within the first six countries of Common Market. The results of a computerized program show differences in location qualities when furnished either with traditional or with nuclear energy systems. In addition to location factor energy some other important factors, as subventions, taxes, labour, and transport costs are analysed, and their influence on industrial location is quantified. (orig.)

  14. The Fort St. Vrain high temperature gas-cooled reactor. Pt. 10

    International Nuclear Information System (INIS)

    In October 1977, during the rise to power test program, the Fort St. Vrain high temperature gas-cooled reactor experienced the first of 37 fluctuation events involving primary coolant outlet temperature, nuclear detector signals, steam generator module gas inlet temperature and steam generator module main and reheat steam temperatures. In a 3 year investigation it was determined that the apparent cause of the fluctuations was movements of core components accompanied by periodic changes in bypass flows and crossflows of primary coolant helium. Installation of region constraint devices has eliminated fluctuations, but a single small primary coolant helium core outlet temperature redistribution is experienced routinely during rise to power. (orig.)

  15. The Fort St. Vrain high temperature gas-cooled reactor. II

    International Nuclear Information System (INIS)

    In field tests in a fossil-fueled facility, performed concurrently with Fort St. Vrain's construction, data indicated that the helium circulator design was well suited to provide primary coolant circulation for the high temperature gas-cooled reactor. After plant installation, primarily during the hot functional tests, a number of time-consuming delays developed caused by cavitation damage on circular speed valves, cavitation and fatigue damage on auxiliary water turbine buckets, water turbine nozzle erosion, static shutdown seal cracks and circulator primary closure helium leakage. After extensive analysis and testing, all of these problems were corrected. Circulators have performed satisfactorily at levels up to 70% of rated power. (Auth.)

  16. Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  17. Design strategies for optically-accessible, high-temperature, high-pressure reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  18. Experimental investigations of actinide release from coated fuel particles for high-temperature reactors

    International Nuclear Information System (INIS)

    The migrational behaviour of actinides in the coated fuel particles proposed for high-temperature reactors is investigated experimentally. Data are described in the framework of the diffusion model. The experimental procedures are presented and the necessary computer codes are discussed. The diffusion coefficients of the actinides - plutonium, americium and curium - as well as of the fission product cesium are derived from the experimental data by a nonlinear least squares fit procedure and are presented in the form of Arrhenius lines D = Do esup(-Q/RT) for U(Th)-O2, HTI-PyC and SiC. (orig.)

  19. Whole-Core Thermal Analysis of Prismatic Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Kim, Min Hwan; Lim, Hong Sik; Jun, Ji Su; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A new method for thermal analysis of prismatic fuel blocks in a very high temperature reactor (VHTR) was developed to overcome the demerits of computational fluid dynamics (CFD) and system calculations. The developed method solves three dimensional heat conduction in prismatic fuel blocks like a CFD code. For the fluid, however, the method adopts one-dimensional conservation equations like a system code. Such a combination enables significantly reduced computational efforts with reasonable computational accuracy. In this paper, the new method has been applied to whole core of PMR200 under full power operating conditions

  20. 2240-MW(th) high-temperature reactor core power density study

    International Nuclear Information System (INIS)

    This study was done to estimate the effects of reducing the design power density of a 2240-MW(t) high-temperature gas-cooled reactor. Core history and thermal hydraulics calculations were performed for average power densities of 5.8 and 7.2 W/cm3 and the use of highly enriched fuel was considered. The fuel temperature conditions for the higher power density were found to be only moderately elevated at normal operating conditions. Economic considerations associated with changes in core performance, core size, and coolant pumping requirements were assessed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-01

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

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

    International Nuclear Information System (INIS)

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

  3. On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  4. Evaluation of high temperature gas cooled reactor performance: Benchmark analysis related to initial testing of the HTTR and HTR-10

    International Nuclear Information System (INIS)

    The Co-ordinated Research Project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance was initiated by the IAEA in 1998 on the recommendation of the Technical Working Group on Gas Cooled Reactors. This CRP was established to foster the sharing of research and associated technical information between participating Member States in the ongoing development of the HTGR as a future source of nuclear energy for high temperature process heat applications and the production of electricity. Of paramount significance in the development of new high temperature gas cooled reactor (HTGR) concepts is the predicted capability for this advanced nuclear plant to achieve a high degree of safety through reliance on passive safety features. Because of this, the investigation and validation of the safety and operational aspects of the HTGR were the primary focus for many of the coordinated research programmes (CRPs) initiated by the IAEA in the 1990s. These included: the neutronic physics behaviour of the HTGR core, fuel performance and fission product behaviour, and the ability of the HTGR to dissipate decay heat by natural transport mechanisms under accident conditions. The principal tools utilized in these CRPs included scientific research and engineering development through analytical evaluation of benchmark problems, application of new and/or existing computer codes and models and utilization of test apparatus and loops for specific component validation. The next important step in bringing this advanced nuclear power programme from concept to actuality is to verify system performance and safety under actual HTGR operating conditions. It is the need for validation via testing in nuclear reactors that was the stimulus for the IAEA to initiate this CRP on 'Evaluation of High Temperature Gas Cooled Reactor Performance'. The principal facilities utilized in the performance of this CRP included Japan's High Temperature Engineering Test Reactor (HTTR), China

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-01

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

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

    International Nuclear Information System (INIS)

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulation routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (''operator'') interaction. The benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses. 3 refs., 3 figs

  7. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  8. Renewable lower reflector structure for a high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The description is given of a renewable lower reflector structure for a high temperature pebble bed reactor of the type comprising a cylindrical or prismatic graphite vessel wrapped in concrete and terminating at its lower end with a conical or pyramidal bottom fitted with a central aperture allowing the pebbles to be discharged by gravity. This structure includes a bed of several layers of protective graphite pebbles on the bottom and, fitted vertically so as to be removable along the centre line of the central aperture through the reflector and the concrete, a graphite block drilled in its centre to allow the discharge of the fuel pebbles and the protective pebbles. The graphite block rises above the level of the central aperture by an extent corresponding to the thickness of the bed when the reactor is working

  9. Transient analysis for the high temperature pebble bed reactor coupled to the energy conversion system

    International Nuclear Information System (INIS)

    This paper describes the results of the calculational coupling between a high temperature reactor code and a thermal hydraulic code for the energy conversion system. This coupling has been developed in order to come to a more detailed and realistic simulation of the entire HTR system. Combining the two codes reduces the number of assumptions that have to be made related to the boundary conditions of the two separate codes. The paper describes the models used for the dynamic components of the energy conversion system, and shows the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system. (author)

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

    International Nuclear Information System (INIS)

    A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 20000F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 15000F could be developed with a high degree of assurance. Process heat at 16000F would require considerably more materials development. While temperatures up to 20000F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR

  11. 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.

  12. A preliminary neutronic evaluation of the high temperature nuclear reactor (HTTR) using reprocessed fuel

    International Nuclear Information System (INIS)

    Highlights: • The HTTR was simulated using reprocessed fuels spiked with thorium and depleted uranium. • The effective neutron multiplication factor and the nuclear fuel evolution during the burn-up were analyzed. • The results indicated that reprocessed fuels can be used in the HTTR. - Abstract: The High Temperature Engineering Test Reactor (HTTR), a 30 MWth, graphite-moderated, helium-cooled reactor constructed by the Japanese government was simulated using reprocessed fuel obtained by UREX+ and spiked with thorium-232 and with depleted uranium. The effective neutron multiplication factor and the nuclear fuel evolution during the burn-up were analyzed. This study was performed using the ORNL SCALE 6.0 code, with CSAS6 and TRITON6 control modules. The results show in a preliminary way that the burn-up of reprocessed fuels in the HTTR core is possible, although the fissile material quantities should be increased while compared with the enrichments of the standard fuel

  13. Appraisal of possible combustion hazards associated with a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The report presents a study of combustion hazards that may be associated with the High Temperature Gas Cooled Reactor (HTGR) in the event of a primary coolant circuit depressurization followed by water or air ingress into the prestressed concrete reactor vessel (PCRV). Reactions between graphite and steam or air produce the combustible gases H2 and/or CO. When these gases are mixed with air in the containment vessel (CV), flammable mixtures may be formed. Various modes of combustion including diffusion or premixed flames and possibly detonation may be exhibited by these mixtures. These combustion processes may create high over-pressure, pressure waves, and very hot gases within the CV and hence may threaten the structural integrity of the CV or damage the instrumentation and control system installations within it. Possible circumstances leading to these hazards and the physical characteristics related to them are delineated and studied in the report

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-21

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

  15. Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

    International Nuclear Information System (INIS)

    Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors. (authors)

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

    International Nuclear Information System (INIS)

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

  17. A small high temperature gas cooled reactor for nuclear marine propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Brugiere, F.; Sillon, C. [Ecole des Applications Militaires de l' Energie Atomique, 50 - Cherbourg (France); Foster, A.; Hamilton, P.; Jewer, S.; Thompson, A.C. [Defence College of Electromechanical Engineering, Nuclear Dept., Military Rd, Gosport (United Kingdom); Kingston, T.; Williams, A.M.; Beeley, P.A. [Rolls-Royce (Marine Power), Raynesway, Derby (United Kingdom)

    2007-07-01

    Results from a design study for a hypothetical nuclear marine propulsion plant are presented. The plant utilizes a small High Temperature Gas Cooled Reactor (HTGCR) similar to the GTHTR300 design by the Japan Atomic Energy Agency with power being generated by a direct cycle gas turbine. The GTHTR300 design is modified in order to achieve the required power of 80 MWth and core lifetime of approximately 10 years. Thermal hydraulic analysis shows that in the event of a complete loss of flow accident the hot channel fuel temperature exceeds the 1600 Celsius degrees limit due to the high power peaking in assemblies adjacent to the inner reflector. Reactor dynamics shows oscillatory behaviour in rapid power transients. An automatic control rod system is suggested to overcome this problem. (authors)

  18. Parameter estimation from dragon high temperature gas cooled reactor dynamic experiments

    International Nuclear Information System (INIS)

    Dynamic experiments were performed on the Dragon high temperature gas cooled reactor at full power, 20 MW. Both terminated ramp and pseudo-random chain code perturbations were applied to a control rod for two amplitudes of reactivity perturbation. Neutron flux and thermocouple signals were observed and recorded together with samples of the inherent noise with the reactor unperturbed. Frequency responses were deduced from the measurements and compared with previous sinusoidal frequency response measurements and theoretical predictions. A simplified model was constructed and optimized by least squares fitting of the equivalent response from the binary cross correlator to the model's output. These optimizations showed that a very simple feedback model is appropriate to Dragon and that a good estimate of the power/reactivity coefficient and temperature coefficient of reactivity may be made. (author)

  19. Energy Neutral Phosphate Fertilizer Production Using High Temperature Reactors: A Philippine Case Study

    International Nuclear Information System (INIS)

    The Philippines may profit from extracting uranium (U) from phosphoric acid during fertilizer production in a way that the recovered U can be beneficiated and taken as raw material for nuclear reactor fuel. Used in a high temperature reactor (HTR) that provides electricity and/or process heat for fertilizer processing and U extraction, energy-neutral fertilizer production, an idea first proposed by Haneklaus et al.,is possible. This paper presents a first case study of the concept regarding a representative phosphate fertilizer plant in the Philippines and exemplary HTR designs (HTR50S and GTHTR300C) developed by the Japan Atomic Energy Agency (JAEA). Three different arrangements (version I-III), ranging from basic electricity supply to overall power supply including on site hydrogen production for ammonia conversion, are introduced and discussed

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

    International Nuclear Information System (INIS)

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

  1. Safety analysis of spent fuel element storage in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Approximately 90000 spent fuel elements will be discharged from a 10 MW high temperature gas-cooled reactor (HTR-10) in its lifetime. The activity of the radioactive fission products in these spent fuel elements will reach 1.0 x 1016 Bq, so these spent fuel elements should be properly managed. HTR-10 spent fuel elements will be discharged into lead-steel containers, with each container designed to receive 2000 fuel elements. These containers will be stored in a concrete compartment inside the reactor building and cooled by air. The author analyzes the release of the radioactive nuclides, the critical safety parameters and the irradiation shielding. The results show that the safety requirements can be met in the HTR-10 spent fuel element storage compartment

  2. Status of international HTGR [high-temperature gas-cooled reactor] development

    International Nuclear Information System (INIS)

    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

  3. Arc-sprayed Coatings for High Temperature Reactors in Titanium Sponge Process

    Science.gov (United States)

    Zhang, Z. L.; Gong, X.; Zhang, N. N.

    2015-07-01

    An arc-sprayed composite coating consisted of an under coat Fe25Cr5Al and a top coat Al5Si was sprayed on the reactor wall surface, which is made of mild steel and used for batch production of titanium sponges. After a trial for 15 production cycles, the morphology of the sprayed coating changed greatly into somewhat like a thermal barrier coating. A large number of composite oxides containing aluminum and chromium had been generated in situ within the protective coatings, especially close to the coating/substrate interface. The oxides generated at the interface could improve the bonding of the coating to the steel substrate together with the surface alumina on the coating surface may provide an effective and long-term protection to the steel substrate of reactor at high temperatures.

  4. Study on transmutation and storage of LLFP using a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    There is a need to temporally store high-level radioactive waste (HLW) until the location of final disposal is decided. HLW contains several types of long-lived fission product (LLFP) which stay radioactive for hundreds of thousands of years. In addition, they tend to be chemically mobile when dissolved into ground water thus may not be suited for geological disposal. A facility that is able to store and incinerate LLFP simultaneously is desirable. The high-temperature gas-cooled reactor (HTGR) is one of the fourth generation nuclear reactors currently under research and it has some favorable characteristics that allow the reactor to destroy LLFP through nuclear transmutation. In this study the capability of HTGR as LLFP transmuter was evaluated in terms of neutron economy. Considering gas turbine high-temperature reactor with 300 MWe nominal capacity (GTHTR300) as HTGR, transmutations of four types of LLFP nuclide were estimated using Monte Carlo transport code MVP and ORIGEN. In addition, burn-up simulations for whole-core region were carried out using MVP-BURN. It was numerically shown that the neutron fluxes change significantly depending on the arrangement of LLFP in the core. When 15 t of LLFP is placed in an ideal manner, the GTHTR300 can sustain sufficient reactivity for one year while transmuting up to 30 kg per year. Additionally, there are more space available for storing larger amount of LLFP without affecting the reactivity. These results suggest that there is a possibility of using GTHTR300 as both LLFP storage and transmuter. (author)

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

    International Nuclear Information System (INIS)

    The High Temperature gas-cooled Reactor (HTR) can fulfil a very useful niche for the purposes of Pu and Minor Actinide (MA) incineration due to its unique and unsurpassed safety features, as well as to the attractive incentives offered by the nature of the coated particle (CP) fuel. No European reactor of this type is currently available, but there has been, and still is, considerable interest internationally. Decisions to construct such a reactor in China and in South Africa have already been made or are about to be made. Apart from the unique and unsurpassed safety features offered by this reactor type, the nature of the CP fuel offers a number of attractive characteristics. In particular, it can withstand burn-ups far beyond that in either LWR or FR systems. Demonstrations as high as 75% FIMA have been achieved. The coated particle itself offers significantly improved proliferation resistance, and finally with a correct choice of the kernel composition, it can be a very effective support for direct geological disposal of the fuel. The overall objective of the PUMA project, a Specific Targeted Research Project (STREP) within the European Union 6th Framework (EU FP6), is to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO2-free energy generation. A number of important issues concerning the use of Pu and MA in gas-cooled reactors have already been dealt with in other projects, or are being treated in ongoing projects, e.g. as part of EU FP6. However, further steps are required to demonstrate the potential of HTRs as Pu/MA transmuters based on realistic/feasible designs of CP Pu/MA fuel and the PUMA focuses on necessary

  6. Advanced Reactor Development in the United States

    International Nuclear Information System (INIS)

    In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE

  7. Design of reactor internals in larger high-temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL)

  8. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950oC. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10-4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10-9 dan 10-8)

  9. Reprocessing of gas turbine high temperature reactor (GTHTR300) spent fuel

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been developing the Gas Turbine High Temperature Reactor (GTHTR300) based on experience gained in development and operations of the High Temperature Engineering Test Reactor (HTTR) in JAERI. The basic fuel cycle concept in Japan is such that all spent fuel shall be reprocessed. Feasibility of the GTHTR300 spent fuel reprocessing was investigated so that the GTHTR300 can comply with the Japanese recycling policy. The Purex process was found to be essentially adaptable except for the head-end treatment. In the head-end process, it was shown that carbon layers and graphite matrix around coated fuel particles are removed from a fuel compact by a burning method, and uranium can be taken out by destruction of the SiC layer with a hard disk crusher, followed by re-burning. Next, the Purex process can be supplied diluted by depleted uranium. To evaluate cost, a preliminary design of the head-end processing plant was studied and reprocessing unit price was evaluated. If the unit cost of waste disposal is assumed nearly equivalent to LWR's, the total fuel cycle cost of GTHTR300 was estimated to be about 1.58 Yen/kWh, which includes the reprocessing cost estimated at about 0.52 Yen/kWh. The economical feasibility of GTHTR300 is thus confirmed. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  10. High temperature CO2 capture using calcium oxide sorbent in a fixed-bed reactor.

    Science.gov (United States)

    Dou, Binlin; Song, Yongchen; Liu, Yingguang; Feng, Cong

    2010-11-15

    The gas-solid reaction and breakthrough curve of CO(2) capture using calcium oxide sorbent at high temperature in a fixed-bed reactor are of great importance, and being influenced by a number of factors makes the characterization and prediction of these a difficult problem. In this study, the operating parameters on reaction between solid sorbent and CO(2) gas at high temperature were investigated. The results of the breakthrough curves showed that calcium oxide sorbent in the fixed-bed reactor was capable of reducing the CO(2) level to near zero level with the steam of 10 vol%, and the sorbent in CaO mixed with MgO of 40 wt% had extremely low capacity for CO(2) capture at 550°C. Calcium oxide sorbent after reaction can be easily regenerated at 900°C by pure N(2) flow. The experimental data were analyzed by shrinking core model, and the results showed reaction rates of both fresh and regeneration sorbents with CO(2) were controlled by a combination of the surface chemical reaction and diffusion of product layer. PMID:20724072

  11. Development plan of high burnup fuel for high temperature gas-cooled reactors in future

    International Nuclear Information System (INIS)

    Plan and status of research and development (R and D) were described on coated fuel particle (CFP) and fuel compacts for the core of small-sized high-temperature gas-cooled reactor (HTGR) HTR50S at second step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce R and D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. The first irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R and D for fuel compact with high packing fraction is planned, because volume fraction of UO2 kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP.We would aim to complete the proof of nuclear/thermal design of second core of HTR50S on integrity of fuel, feeding back results of R and Ds and nuclear/thermal design to each other. In addition, we describe outline of R and D plans for core of HTR50S in phase II and practical HTGR in Japan in future, naturally safe HTGR. (author)

  12. Evaluation method and prediction result of fuel behavior during the High Temperature Engineering Test Reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Yoshimuta, Shigeharu; Sato, Masashi; Saito, Kenji; Tobita, Tsutomu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    Small amounts of additional failure of HTGR (High Temperature Gas-cooled Reactor) fuel will occur during operation. In the safety design requirements for the High Temperature Engineering Test Reactor (HTTR) fuel, the additional failure fraction in the coating layers of the coated fuel particles is limited less than 0.2% through the full service period. The failure fraction should be know during the HTTR operation. Short-lived fission gases are released from the through-coatings-failed particles and contamination uranium in the fuel compact matrix since the coating layers can retain short-lived fission gases. Then fission gas concentration in the primary coolant reflects the failure fraction in the core. Based on fuel fabrication data (exposed uranium fractions and the SiC failure fractions) and the HTTR operating condition, the though-coatings-failure fraction and release fraction of {sup 88}Kr are analytically predicted. The results are as follows. (1) The intact particles will not fail by kernel migration, Pd-SiC corrosion and internal pressure, however, some of the as-fabricated SiC-failed particles will be the through-coatings-failed particles by the pressure vessel failure. (2) The release fraction of {sup 88}Kr, that will be determined mainly by the release from the contamination uranium in the fuel compact matrix, will be less than 10{sup -6} considering the additional through-coatings-failure fraction. (author)

  13. High temperature CO2 capture using calcium oxide sorbent in a fixed-bed reactor

    International Nuclear Information System (INIS)

    The gas-solid reaction and breakthrough curve of CO2 capture using calcium oxide sorbent at high temperature in a fixed-bed reactor are of great importance, and being influenced by a number of factors makes the characterization and prediction of these a difficult problem. In this study, the operating parameters on reaction between solid sorbent and CO2 gas at high temperature were investigated. The results of the breakthrough curves showed that calcium oxide sorbent in the fixed-bed reactor was capable of reducing the CO2 level to near zero level with the steam of 10 vol%, and the sorbent in CaO mixed with MgO of 40 wt% had extremely low capacity for CO2 capture at 550 deg. C. Calcium oxide sorbent after reaction can be easily regenerated at 900 deg. C by pure N2 flow. The experimental data were analyzed by shrinking core model, and the results showed reaction rates of both fresh and regeneration sorbents with CO2 were controlled by a combination of the surface chemical reaction and diffusion of product layer.

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

    International Nuclear Information System (INIS)

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

  15. Development of high temperature fission counter-chamber(FC)S for a top entry loop type fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype high temperature fission counter-chambers have been made as neutron detectors for installation in the reactor vessel of the 600MWe-class top entry loop type fast breeder reactor. Using these prototypes as samples, a high-temperature endurance test has been conducted. The validity of the prototypes has been established by the test results, which show that the prototypes nearly satisfy the design performance. (author)

  16. Near term test plan using HTTR (high temperature engineering test reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Shoji, E-mail: takada.shoji@jaea.go.jp [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan); Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Yanagi, Shunki [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) (Japan); Nishihara, Tetsuo [Policy Department and Administration Department, JAEA (Japan); Fukaya, Yuji [HTGR Design Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Goto, Minoru [HTGR Safety Evaluation Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Tachibana, Yukio [HTGR Design Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Sawa, Kazuhiro [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) (Japan)

    2014-05-01

    JAEA has carried out research and development to establish the technical basis of high temperature gas cooled reactors (HTGRs) using HTTR. In order to connect hydrogen production system to HTTR, it is necessary to ensure the stability of plant dynamics when the thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the stable reactor dynamics and to gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted by HTTR to confirm the validity of the calculated temperature coefficient of reactivity. A loss of forced cooling (LOFC) test using HTTR has been carried out to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled.

  17. Behavior of a high-temperature gas reactor with transuranic fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fortini, A.; Pereira, C.; Sousa, R.V.; Veloso, M.A.F.; Costa, A.L.; Silva, C.A.; Cardoso, F.S., E-mail: fortini@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    In this work, we modeled a high-temperature gas reactor, HTGR, of prismatic block type using the SCALE 6.0 code to analyze the use of transuranic fuel in these reactors. To represent the concept, the Japanese HTTR reactor was chosen. The fuels considered used transuranic elements from UREX+ reprocessing of burned PWR fuel spiked with depleted U or Th. The calculations, performed for typical temperatures of HTR reactors, showed that, in mixtures with the same percentage of fissile material, the initial effective multiplication factor, K{sub eff} , is higher in the mixtures containing Th than that with U. Comparisons between the two types of fuel were performed using fuel pairs with the same initial K{sub eff}. During burn-up, the two mixtures show a slow and practically equal decrease in K{sub eff}. For the same level of burnup, mixtures containing Th show greater effectiveness in burning transuranics and total plutonium when compared to corresponding mixtures with depleted U. (author)

  18. Present status of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

  19. Considerations on the design of a helium circulator for a high temperature modular reactor system

    International Nuclear Information System (INIS)

    A modular helium cooled, high temperature reactor system with a thermal output of 200 MW per reactor has been developed by the KWU group for cogeneration of electricity and process steam. The flow of the reactor coolant - Helium at 60 bars and 250/700 deg. C is maintained by one circulator per reactor. The circulator is driven by a variable speed Siemens asynchronous motor and is submerged in the helium primary system. For operational reasons high reliability and availability of the circulator is required. The operational requirements for the circulator design are presented in this paper. The actual design has been carried out in close cooperation with the designer and manufacturer of all submerged circulators operating in AGR plants in Great Britain, James Howden Co. Renfrew, Scotland. Design solutions received so far and mainly based on sufficiently proven components - such as oil bath lubricated bearing systems - will be described. Special attention will be paid on the necessary test work; especially for the prototype to confirm the lay out. (author). 9 figs

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

    International Nuclear Information System (INIS)

    In 1969 JAERI started the design study of the Experimental Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR), and trial design, preliminary design, conceptual design, comprehensive system design and the first and second stage of detailed design have been carried out. Hereafter JAERI is going to pursue the rationalized Experimental VHTR system which maintains the required functions and performance and has the potential for reducing the construction cost, utilizing extensively the inherent safety features of HTGRs. In the current design, i.e. the second stage of detailed design, the reactor outlet coolant temperature is 9500C to aim earlier construction of the Experimental VHTR, according to the specification in ''Long-term plan for the development and utilization of nuclear energy'' revised by Japan Atomic Energy Commission in June 1982. This report presents the results based mainly on the comprehensive system design (completed by 1980.3) which is the last overall system design of the Experimental VHTR aiming 10000C reactor outlet coolant temperature and partially on the first stage (completed by 1981.3) of detailed design in the form of ''an application of reactor construction permit, Appendix 8'', excepting comformance with ''Safety Design Requirements'' which correspond to ''Safety Design Criteria for Water Cooled Nuclear Power Plants issued by Japan Nuclear Safety Commission''. (author)

  1. High-temperature gas-cooled-reactor steam-methane reformer design

    International Nuclear Information System (INIS)

    The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam reforming reaction, is currently being evaluated as an energy source/application for use early in the 21st century. The steam-methane reforming reaction is an endothermic reaction at temperatures approximately 7000C and higher, which produces hydrogen, carbon monoxide and carbon dioxide. The heat of the reaction products can then be released, after being pumped to industrial site users, in a methanation process producing superheated steam and methane which is then returned to the reactor plant site. In this application the steam reforming reaction temperatures are produced by the heat energy from the core of the HTGR through forced convection of the primary or secondary helium circuit to the catalytic chemical reactor (steam reformer). This paper summarizes the design of a helium heated steam reformer utilized in conjunction with a 1170 MW(t) intermediate loop, 8500C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, materials selection and the structural design analysis

  2. High temperature corrosion of structural materials under gas-cooled reactor helium

    International Nuclear Information System (INIS)

    The Generation IV International Forum has selected six promising nuclear power systems for further collaborative investigations and development. Among these six concepts, two candidates are Gas Cooled Reactors (GCR), namely the Very High Temperature Reactor (VHTR) and the Gas-cooled Fast Reactor (GFR). The CEA has launched a R and D program on the metallic materials for application in an innovative GCR. Structural GCR alloys have been extensively studied in the past three decades. Some critical aspects for the steels and nickel base alloys resistance under the service conditions are microstructural stability, creep strength and compatibility with the cooling gas. The coolant, namely helium, proved to contain impurities mainly H2, CO, CH4, N2 and steam in the microbar range that interact with metals at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the atmosphere characteristics, primarily the effective oxygen partial pressure and carbon activity, on the temperature and on the alloys chemical composition. These structural transformations can notably influence the mechanical properties: carburisation may induce a loss in toughness and ductility whereas decarburisation impedes the creep strength. There is a valuable theoretical as well as practical knowledge on the corrosion of high temperature alloys in the primary circuit of a GCR but this past experience is not sufficient to qualify every component in a future reactor. On the one hand, the material environment could be significantly different from the former GCR's, especially regarding the higher temperature. On the other hand, the materials of interest are partly different. Ni-Cr-W alloys, for instance, may offer significant improvement in the maximum operating temperature as far as the mechanical properties are concerned. However, their corrosion resistance toward the GCR atmosphere is still unknown. We describe here our first corrosion tests of Haynes

  3. Concept of a high temperature integrated multi-modular thermal reactor

    International Nuclear Information System (INIS)

    Highlights: ► Concept high temperature integrated multi-modular reactor core. ► BeO–UO2 fueled, ScCO2 cooled, with Brayton cycle turbo-machinery. ► Critical configuration achieved with several independent sub-critical modules. ► Each module fully contains the heat supply and balance of plant systems. ► Customizable plant configuration with a seven module, 10 MWth design presented. - Abstract: The high temperature integrated multi-modular thermal reactor is a first-of-a-kind small modular reactor that uses an enhanced conductivity BeO–UO2 fuel with supercritical CO2 coolant to drive turbo-machinery in a direct Brayton cycle. The core consists of several self-contained pressurized modules, each containing fuel elements in pressurized channels surrounded by a graphite moderator, and Brayton cycle turbo-machinery. Each module is subcritical by itself, and when several modules are brought into proximity of one another, a single critical core is formed. The multi-modular approach and use of BeO–UO2 fuel with graphite moderator and supercritical CO2 coolant leads to an inherently safe system capable of high efficiency operation. The pressure channel design and multi-modular approach eliminates engineering challenges associated with large pressure vessels, and the subcriticality of the modules ensures inherent safety during construction, transportation, and after decommissioning. A feasible configuration consisting of seven modules operating for 14.7 years at 10 MWth power using 5% BeO and 5% 235U enriched fuel is presented in this paper. The results of the conceptual analysis include flux and power distributions as well as potential power shaping and reactivity coefficients were analyzed

  4. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  5. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

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

    International Nuclear Information System (INIS)

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

  7. Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress

    International Nuclear Information System (INIS)

    A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (∼160 sec) would be significantly earlier than the previous predictions (∼150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected

  8. Efficiency Testing of the Air Cleaning System for a High Temperature Reactor

    International Nuclear Information System (INIS)

    The Los Alamos Ultra High Temperature Reactor Experiment (UHTREX) utilizes a helium-cooled, graphite-moderated reactor, employing refractory fuel elements. Under accident conditions, the effluent that may be released from this reactor requires an air-cleaning system capable of reducing radioactive gas and particulate contaminants to safe levels. Dioctyl phthalate and iodine-131 were used as test aerosols for the HEPA and activated carbon filters, respectively. Methods of aerosol generation and test procedures are detailed for the preinstallation tests of the carbon and in-place testing of the carbon and HEPA filters. The importance of visual inspection of the HEPA filters prior to installation and supervision of filter installation is discussed. In-place tests indicated desirable design changes which would (1) simplify in-place testing procedures, (2) expedite installation and future changing of the filters, and (3) ensure operation of a more efficient system. Problems encountered during in-place testing, recommendations for the design of similar systems, and acceptance criteria used at LASL are discussed. (author)

  9. ECUT energy data reference series: high-temperature materials for advanced heat engines

    Energy Technology Data Exchange (ETDEWEB)

    Abarcar, R.B.; Hane, G.J.; Johnson, D.R.

    1984-07-01

    Information that describes the use of high-temperature materials in advanced heat engines for ground transportation applications is summarized. Applications discussed are: automobiles, light trucks, and medium and heavy trucks. The information provided on each of these modes includes descriptions of the average conversion efficiency of the engine, the capital stock, the amount of energy used, and the activity level as measured in ton-miles.

  10. Production of liquid fuels with a high-temperature gas-cooled reactor

    Science.gov (United States)

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

    An exploration is made of the technical, economic and environmental impact feasibility of integrating coal liquefaction methods directly and indirectly with a nuclear reactor source of process heat, with stress on the production of synthetic jet fuel. Production figures and operating costs are compared for indirect conventional and nuclear processes using Lurgi-Fischer-Tropsch technology with direct conventional and nuclear techniques employing the advanced SRC-II technology, and it is concluded that significant advantages in coal savings and environmental impact can be expected from nuclear reactor integration.

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

    International Nuclear Information System (INIS)

    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

  12. Reactor pressure vessel design of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The reactor pressure vessel (RPV) of the HTTR is 5.5m (inside diameter), 13.2m (inside height), and 122mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr-1Mo steel is chosen as the material for RPV. The temperature reaches about 400 deg. C at normal operation. The fluence of the RPV is estimated to be less than 1 x 1017n/cm2 (E > 1MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr-1Mo steel for the RPV; embrittlement parameters, J-factor and X-bar are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr-1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and X-bar , and nil-ductility transition temperatures, TNDT, by drop weight tests, are shown. In-service inspection and results of R and Ds are also described

  13. Current status and future development of modular high temperature gas cooled reactor technology

    International Nuclear Information System (INIS)

    This report includes an examination of the international activities with regard to the development of the modular HTGR coupled to a gas turbine. The most significant of these gas turbine programmes include the pebble bed modular reactor (PBMR) being designed by ESKOM of South Africa and British Nuclear Fuels plc. (BNFL) of the United Kingdom, and the gas turbine-modular helium reactor (GT-MHR) by a consortium of General Atomics of the United States of America, MINATOM of the Russian Federation, Framatome of France and Fuji Electric of Japan. Details of the design, economics and plans for these plants are provided in Chapters 3 and 4, respectively. Test reactors to evaluate the safety and general performance of the HTGR and to support research and development activities including electricity generation via the gas turbine and validation of high temperature process heat applications are being commissioned in Japan and China. Construction of the high temperature engineering test reactor (HTTR) by the Japan Atomic Energy Research Institute (JAERI) at its Oarai Research Establishment has been completed with the plant currently in the low power physics testing phase of commissioning. Construction of the high temperature reactor (HTR-10) by the Institute of Nuclear Energy Technology (INET) in Beijing, China, is nearly complete with initial criticality expected in 2000. Chapter 5 provides a discussion of purpose, status and testing programmes for these two plants. In addition to the activities related to the above mentioned plants, Member States of the IWGGCR continue to support research associated with HTGR safety and performance as well as development of alternative designs for commercial applications. These activities are being addressed by national energy institutes and, in some projects, private industry, within China, France, Germany, Indonesia, Japan, the Netherlands, the Russian Federation, South Africa, United Kingdom and the USA. Chapter 6 includes details

  14. Technical assessment of gas turbine cycle for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The gas turbine cycle appears to be the best near-term power conversion method for the high temperature gas-cooled reactor (HTGR). The author extensively investigates the gas turbine cycle including direct cycle, open indirect cycle and closed indirect cycle with medium of helium, nitrogen and air. Each cycle is analyzed and optimized from the thermodynamic standpoint and its turbo-machine is aerodynamically designed. As a result, the direct cycle with helium is an ideal option for the HTGR gas turbine cycle; however it is not easy to be realized based on current technology. The closed indirect cycle with helium or nitrogen is a practical one at present time, which can get the gas turbine cycle and lay technical bases for the future direct cycle

  15. Quantitative analysis of strontium 89 and strontium 90 used for experiments with high temperature reactors

    International Nuclear Information System (INIS)

    A summary is given about hitherto studied and applied version of sample pretreatment and separation techniques of 89Sr and 90Sr from fission and activation products in plate-out experiments for security studies of graphite moderated gas cooled high temperature pebble bed reactors. The latest and most succesfully applied version with a combined double-column system is described and its performance and results are discussed in detail. Up to 50 samples can be analyzed weekly. The experimental (primary) data are evaluated by the use of a computer program. The reliability is between 4-30%, the accuracy is better then 10%. The sensitivity is better than 5.10-11Ci. Decontamination factors better then 103 are achieved. (T.G.)

  16. Unirradiated high temperature reactor fuel element head-end reprocessing tests

    International Nuclear Information System (INIS)

    For several years, the United States and the Federal Republic of Germany (FRG) have engaged in a successful cooperative program to develop high temperature gas-cooled reactor (HTGR) fuel cycle technology. Recent tests in reprocessing pilot plant facilities at General Atomic Company have demonstrated the feasibility of performing HTGR head-end unit operations for both spherical (German) and block-type (American) fuel elements in a single process line. Because of an unexpected high fines generation and elutriation rate, extended fluidized bed primary burning of FRG fuel material was impossible to accomplish with the burner system and operating procedures optimized for U.S. fuel burning. Operational modification, including startup with a carbon-poor bed and reduction of the fluidizing velocity, resulted in dramatic improvements in FRG fuel-burning behavior and allowed extended processing campaigns. Additional modifications to the fines recycle system and burner are recommended to optimize the system for processing of FRG fuels

  17. High temperature blankets for non-electrical/electrical applications of fusion reactors: Annual report, [1983

    International Nuclear Information System (INIS)

    During FY '83 the Li2O solid-breeder, helium-cooled canister blanket emerged as the LLNL-UW choice for driving the low-temperature (2, high-temperature outer zone for driving the GA hydrogen synfuel process. Providing 3-dimensional neutronics analysis of power deposition and tritium breeding in both blankets was an important part of the UW-Rowe and Assoc. work. In both the LLNL-UW and MARS studies, the fusion driver as the Axi-Cell, A-cell version of the tandem mirror reactor (TMR). Physics parameters consistent with the synfuel interface were determined as part of the work. Defining and analyzing the thermal-electric interfaces between the TMR and the synfuel process continues to be of prime importance. The analysis of thermal transport and energy conversion in the interface, as well as thermal hydraulics analysis of the blanket, were part of the UW-Rowe Assoc. work

  18. The high temperature reactor - an important tool in meeting the challenge of world energy supply

    International Nuclear Information System (INIS)

    A growing and, in its majority, poor mankind will need increasing amounts of energy at moderate prices. At the same time, ecological stresses on our environment, on the forests of the Third World (firewood crisis), and on the climate must be limited. The High Temperature Reactor (HTR) is a well-suited answer to all challenges, as it can supply electricity safely and economically, be built close to process steam and district heat consumers, procure more hydrocarbons from coal relative to a given release of CO2, and has the potential of splitting water with high efficiency. At times of effluent fossil fuels, however, and not yet apparent need to restrict their use for reasons of climate, individual companies cannot bear the development and introduction of HTRs all by themselves. Therefore governments are called upon for support. (author)

  19. Helium sampling and analyzing system of 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The helium purification system of 10 MW high temperature gas-cooled reactor consists of the purification equipments and their accessories. The purification equipments include a copper oxide bed, a molecular sieve absorber, a low temperature absorbed, and etc. The gas sampling and analyzing system is consisted of the gas chromatograph, moisture probe, and infrared analyzing instrument. The moisture probe and infrared analyzing instrument both conform to the design requirement, and consecutive inspection of H2O, CO, CO2 can be carried out for the primary helium circuit. The gas chromatograph can also meet the design requirement, and so the intermittent sampling and analyzing of H2, O2, N2, CH4, CO and CO2 can be carried out for the primary helium circuit. (authors)

  20. Probabilistic method for evaluation of reactivity margin of experimental very high temperature reactor

    International Nuclear Information System (INIS)

    A probabilistic method is proposed to evaluate in the core desigh stage the possibility that the safety criteria in reactivity margin are satisfied, taking into consideration the uncertainties in design calculation. In application of the method to design study cores of Experimental Very High Temperature Reactor, investigations are made for the relation between the design accuracy and the probability that the safety criteria in both shut-down and operation margins are satisfied. In conclusion, for the MARK-III core, with the correlation disregarded, the ratio of the standard deviations to the design values must be less than 0.79 and 5.3% respectively for the cold clean effective multiplication factor and the reactivity worths of control rods, burnable poisons and core temperature rise, in order that the probability is larger than 99.7% (three times the sigma limit). With the correlation regarded, the ratios must be considerably smaller. (author)

  1. An inspection standard of fuel for the high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author).

  2. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  3. Helium gas turbine plant in Oberhausen - a milestone for the development of high-temperature reactors

    International Nuclear Information System (INIS)

    The 50 MW helium turbine from the GHH Sterkrade for the new thermal power plant of Energy Supply, Oberhausen (EVO), works according to the method of closed cycles proposed by both professors Ackeret and Keller in 1939. In this closed gas turbine method, the working medium does not take part in the combustion, but is heated from the outside in a superheater. After the heated medium has given off its energy, it must be cooled down again in an after-cooler so that the power will not be too great when newly compressing the circulating medium. A future field of application of the gas turbine in closed cycle is the direct coupling with helium cooled high temperature reactors for over 1,000 MW electric power. Layout and construction of the turbine plant are described. (orig./LH)

  4. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  5. Applications for a high temperature gas cooled nuclear reactor in oil shale processing

    International Nuclear Information System (INIS)

    Results are presented of a study concerning possible applications for a high temperature gas cooled reactor as a process heat source in oil shale retorting and upgrading. Both surface and in situ technologies were evaluated with respect to the applicability and potential benefits of introducing an outside heat source. The primary focus of the study was to determine the fossil resource which might be conserved, or freed for higher uses than furnishing process heat. In addition to evaluating single technologies, a centralized upgrading plant, which would hydrotreat the product from a 400,000 bbl/day regional shale oil industry was also evaluated. The process heat required for hydrogen manufacture via steam reforming, and for whole shale oil hydrotreating would be supplied by an HTGR. Process heat would be supplied where applicable, and electrical power would be generated for the entire industry

  6. Sorption of caesium and strontium by graphite materials in gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    The experiments have revealed that coked phenol resin binder has got an extremely high sorption capacity for Cs and Sr. For this reason the sorption capacity of A3 matrix graphite for fuel elements exceeds the capacity of the highly graphitized material that does not contain this component. The strong chemical binding for Cs and Sr by chemisorption indicates a retention of these elements when the nucleus is heated up by accident. The release calculations carried out with definite sorption isotherms revealed a larger retention effect by sorption for Sr than for Cs. In this respect the matrix graphite in the ball-shaped fuel elements is of special importance for the retention. It is applied at German high temperature reactors and contains non-graphitized phenol resin binder. (orig./DG)

  7. Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Highlights: • We developed an adequate model for the identification of damage domains of the HTTR. • We analysed an anticipated operational transient, using the HTTR5+/GASTEMP code. • We simulated several transients of the same sequence. • We identified the corresponding damage domains using two methods. • We calculated exceedance frequency using the two methods. - Abstract: This paper presents an assessment analysis of damage domains of the 30 MWth prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA

  8. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    International Nuclear Information System (INIS)

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data

  9. Spectral emissivity measurements of candidate materials for very high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cao, G.; Weber, S.J.; Martin, S.O.; Anderson, M.H. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Sridharan, K., E-mail: kumars@cae.wisc.edu [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States)

    2012-10-15

    Heat dissipation by radiation is an important consideration in VHTR components, particularly the reactor pressure vessel (RPV), because of the fourth power temperature dependence of radiated heat. Since emissivity is the material property that dictates the ability to radiate heat, measurements of emissivities of materials that are being specifically considered for the construction of VHTR become important. Emissivity is a surface phenomenon and therefore compositional, structural, and topographical changes that occur at the surfaces of these materials as a result of their interactions with the environment at high temperatures will alter their emissivities. With this background, an experimental system for the measurement of spectral emissivity has been designed and constructed. The system has been calibrated in conformance with U.S. DoE quality assurance standards using inert ceramic materials, boron nitride, silicon carbide, and aluminum oxide. The results of high temperature emissivity measurements of potential VHTR materials such as ferritic steels SA 508, T22, T91 and austenitic alloys IN 800H, Haynes 230, IN 617, and 316 stainless steel have been presented.

  10. Spectral emissivity measurements of candidate materials for very high temperature reactors

    International Nuclear Information System (INIS)

    Heat dissipation by radiation is an important consideration in VHTR components, particularly the reactor pressure vessel (RPV), because of the fourth power temperature dependence of radiated heat. Since emissivity is the material property that dictates the ability to radiate heat, measurements of emissivities of materials that are being specifically considered for the construction of VHTR become important. Emissivity is a surface phenomenon and therefore compositional, structural, and topographical changes that occur at the surfaces of these materials as a result of their interactions with the environment at high temperatures will alter their emissivities. With this background, an experimental system for the measurement of spectral emissivity has been designed and constructed. The system has been calibrated in conformance with U.S. DoE quality assurance standards using inert ceramic materials, boron nitride, silicon carbide, and aluminum oxide. The results of high temperature emissivity measurements of potential VHTR materials such as ferritic steels SA 508, T22, T91 and austenitic alloys IN 800H, Haynes 230, IN 617, and 316 stainless steel have been presented.

  11. In Situ Measurements of Spectral Emissivity of Materials for Very High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    G. Cao; S. J. Weber; S. O. Martin; T. L. Malaney; S. R. Slattery; M. H. Anderson; K. Sridharan; T. R. Allen

    2011-08-01

    An experimental facility for in situ measurements of high-temperature spectral emissivity of materials in environments of interest to the gas-cooled very high temperature reactor (VHTR) has been developed. The facility is capable of measuring emissivities of seven materials in a single experiment, thereby enhancing the accuracy in measurements due to even minor systemic variations in temperatures and environments. The system consists of a cylindrical silicon carbide (SiC) block with seven sample cavities and a deep blackbody cavity, a detailed optical system, and a Fourier transform infrared spectrometer. The reliability of the facility has been confirmed by comparing measured spectral emissivities of SiC, boron nitride, and alumina (Al2O3) at 600 C against those reported in literature. The spectral emissivities of two candidate alloys for VHTR, INCONEL{reg_sign} alloy 617 (INCONEL is a registered trademark of the Special Metals Corporation group of companies) and SA508 steel, in air environment at 700 C were measured.

  12. Chemical analysis of nickel- and iron-base high-temperature alloys for nuclear reactor

    International Nuclear Information System (INIS)

    The Committee studied problems in analysis of alloys used for High-Temperature Gas Cooled Reactor from September 1970 to February 1976. The alloys selected from the standpoint of analytical chemistry are Inconel 600, Incoloy 800, Inconel X750, Inco 713C and Hastelloy X. Nine standard samples (JAERI-R 1 to JAERI-R 9) of the high-temperature alloys were prepared primarily for X-ray fluorescence method. Eighteen research institutions in Japan participated in cooperative analyses of the standard samples for 19 elements (C, Si, Mn, P, S, Ni, Cr, Fe, Mo, Cu, W, V, Co, Ti, Al, B, Nb, Ta, Zr). Prior to analyses of the standard samples, 8 cooperative samples (A-H) were analyzed to develop and evaluate analytical methods. Described in this report are preparation and their characteristics of the standard samples, results of analyses, and 93 analytical methods. The results of the cooperative experiments on atomic absorption spectrophotometry and X-ray fluorescence method are also described. (auth.)

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

    International Nuclear Information System (INIS)

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

  14. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  15. Determination of an instability temperature for alloys in the cooling gas of a high temperature reactor

    International Nuclear Information System (INIS)

    High temperature alloys designed to be used for components in the primary circuit of a helium cooled high temperature nuclear reactor show massive CO production above a certain temperature, called the instability temperature T/sub i/, which increases with increasing partial pressure of CO in the cooling gas. At p/sub CO/ = 15 microbar, T/sub i/ lies between 900 and 950 degrees C for the four alloys under investigation: T/sub i/ is lowest for the iron base alloy Incoloy 800 H and increases for the nickel base alloys in the order Inconel 617, HDA 230 and Nimonic 86. Measurements of T/sub i/ made at 3 different laboratories were compared and shown to agree for p/sub CO/25 microbar, compatible with CO production by a reaction of Cr2O3 with carbides. Some measurements of T/sub i/ on HDA 230 and Nimonic 86 were performed in the course of simulated reactor disturbances. They showed that the oxide layer looses its protective properties above T/sub i/. A highlight of the examinations was the detection of eta-carbides (M6C) with unusual properties. M6C is the only type of carbide occuring in HDA 230. An eta-carbide with a lattice constant of 1088.8 pm had developed at the surface of Nimonic 86 during pre-oxidation before the disturbance simulation. Its composition is estimated at Ni3SiMo2C. Eta-carbides containing Si and especially eta-carbides with lattice constants as low as 1088.8 pm have been described only rarely until now. (author)

  16. Approaches to experimental validation of high-temperature gas-cooled reactor components

    International Nuclear Information System (INIS)

    Highlights: ► Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. ► Vibroacoustic investigations. ► Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. ► Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 °C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor electromagnetic bearings system operability.

  17. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  18. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  19. High-temperature reactors for underground liquid-fuels production with direct carbon sequestration

    International Nuclear Information System (INIS)

    The world faces two major challenges: (1) reducing dependence on oil from unstable parts of the world and (2) minimizing greenhouse gas emissions. Oil provides 39% of the energy needs of the United States, and oil refineries consume over 7% of the total energy. The world is running out of light crude oil and is increasingly using heavier fossil feedstocks such as heavy oils, tar sands, oil shale, and coal for the production of liquid fuels (gasoline, diesel, and jet fuel). With heavier feedstocks, more energy is needed to convert the feedstocks into liquid fuels. In the extreme case of coal liquefaction, the energy consumed in the liquefaction process is almost twice the energy value of the liquid fuel. This trend implies large increases in carbon dioxide releases per liter of liquid transport fuel that is produced. It is proposed that high-temperature nuclear heat be used to refine hydrocarbon feedstocks (heavy oil, tar sands, oil shale, and coal) 'in situ ', i.e., underground. Using these resources for liquid fuel production would potentially enable the United States to become an exporter of oil while sequestering carbon from the refining process underground as carbon. This option has become potentially viable because of three technical developments: precision drilling, underground isolation of geological formations with freeze walls, and the understanding that the slow heating of heavy hydrocarbons (versus fast heating) increases the yield of light oils while producing a high-carbon solid residue. Required peak reactor temperatures are near 700 deg. C-temperatures within the current capabilities of high-temperature reactors. (authors)

  20. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MWth reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs

  1. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  2. Physical-property study on liquid fluoride-salt-cooled high temperature reactor loaded with different kinds of fuel mixtures

    International Nuclear Information System (INIS)

    Background: Liquid fluoride-salt-cooled high temperature reactor, one of the GEN-IV reactors, has great advantages comparatively, and the research on its fuel is of great significance. Purpose: The aim is to study the physical properties of high-temperature pebble-bed reactors loaded with six different fuel mixtures. Methods: We used SCALE5.1 package to compute some important parameters like excess reactivity, full power operation days, burnup and neutron spectrum. Results: The results show that the true conversions of 232Th are much less than those of 238U when mixed with 233U or 235U. With 239Pu for starting, 232Th contributes to making the reactor running longer than that with 238U. Conclusion: With respect to saving fuel and extending the life of the core, without online refueling and reprocessing, 232Th is under performed in thermal reactor compared with 238U, however, it is the opposite in epithermal neutron reactor. (authors)

  3. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

  4. An improved porous media approach to thermal–hydraulics analysis of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Highlights: • The capabilities of Thermo Hydraulic Porous Program (THPP) for simulating pebble bed and block fuel elements are demonstrated. • Outputs of the comparisons between THPP and the well-established codes demonstrates good agreement concerning steady state operation conditions. • Comparisons of results between THPP and the thermal-hydraulics codes have shown good agreement for simulating pebble type fuel reactors. • Results obtained from THPP show good agreement to simulate the block type of reactor. - Abstract: A precise thermal–hydraulics model is of great importance for developing more effective designs of High Temperature Gas Cooled Reactors (HTGR). Recently, several advancements have been made in the methods of analysis of porous media which could be of significant value in the development of more precise and robust codes. The objective of this research is to incorporate some of the most recent improvements in the development of a new 2D program for thermal–hydraulics analysis of modular high temperature reactors. The program is mainly based on the solution of a coupled set of mass, energy and momentum conservation equations for the gas flow, along with the energy conservation equation in the solid. The energy conservation has been cast for non-equilibrium conditions. A suitable implementation could enable the program to handle both well-known types of HTGR, namely the pebble bed and prismatic. To this aim, an appropriate set of constitutive equations for effective heat conductivity of solid, pressure drop, and heat transfer coefficient were used for each reactor type. One should be aware of the specific case of effective heat conductivity according to its importance in the analysis where its dependence on temperature and dose should be considered. Moreover, two distinct models have been adapted for a better estimation of effective heat conductivity. The finite-volume method has been used for numerical solution of the conservation

  5. Crystal plasticity based finite element model for simulation of high temperature deformation behavior of Niobium based alloys for high temperature reactors

    International Nuclear Information System (INIS)

    For structural components of compact high temperature reactors, Niobium based alloys are some of the candidate materials which are being studied extensively by various researchers. These alloys have excellent high temperature mechanical properties for temperature range as high as 1000 to 1300 deg. C. The NbZrC alloys form different types of carbides which impart high temperature strength to these alloys. The alloy also possesses good ductility at elevated temperatures. In order to understand the material deformation behavior of the alloy, a crystal plasticity based model has been used in simulation of material stress-strain curve at various elevated temperatures. It is very important to take into account of the underlying microstructure of the material in order to develop a reliable constitutive model for predicting the elevated temperature strength of these alloys. Crystal plasticity based models are suitable for this purpose as these take into account of the crystal orientations of different grains as well as the effect of various microstructural features on the onset of plasticity and plastic hardening mechanisms in these materials. However, it is computationally expensive to incorporate the explicit models of different features of the microstructure in a crystal plasticity based framework to simulate the response of the polycrystalline micro-structure of these alloys. The aim of this work is to develop a physically motivated multi-scale approach for simulation of response of these types of alloys. At the lower scale, i.e., at the grain level, the crystal plasticity model simulates the response of various types of microstructures (with different morphology of precipitates) within a single crystal. The microstructures are designed with various shapes and volume fractions of precipitates. The lower scale model is homogenized as a function of various microstructural parameters and the homogenized model is used at the polycrystalline level of crystal plasticity

  6. Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2009-09-01

    A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen

  7. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    International Nuclear Information System (INIS)

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies

  8. Application of the hybrid diffusion-transport spatial homogenization method to a high temperature test reactor benchmark problem

    International Nuclear Information System (INIS)

    The recently developed Hybrid Diffusion-Transport Spatial Homogenization (DTH) Method was previously tested on a benchmark problem typical of a boiling water reactor. In this paper, the DTH method is tested in a 1-D benchmark problem based on the Japanese High Temperature Test Reactor (HTTR). This acts as a verification of the method for a reactor that is optically thinner than the original BWR test benchmark. (author)

  9. Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

    International Nuclear Information System (INIS)

    A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

  10. Proceedings of the national symposium on materials and processing: functional glass/glass-ceramics, advanced ceramics and high temperature materials

    International Nuclear Information System (INIS)

    With the development of materials science it is becoming increasingly important to process some novel materials in the area of glass, advanced ceramics and high temperature metals/alloys, which play an important role in the realization of many new technologies. Such applications demand materials with tailored specifications. Glasses and glass-ceramics find exotic applications in areas like radioactive waste storage, optical communication, zero thermal expansion coefficient telescopic mirrors, human safety gadgets (radiation resistance windows, bullet proof apparels, heat resistance components etc), biomedical (implants, hyperthermia treatment, bone cement, bone grafting etc). Advanced ceramic materials have been beneficial in biomedical applications due to their strength, biocompatibility and wear resistance. Non-oxide ceramics such as carbides, borides, silicides, their composites, refractory metals and alloys are useful as structural and control rod components in high temperature fission/ fusion reactors. Over the years a number of novel processing techniques like selective laser melting, microwave heating, nano-ceramic processing etc have emerged. A detailed understanding of the various aspects of synthesis, processing and characterization of these materials provides the base for development of novel technologies for different applications. Keeping this in mind and realizing the need for taking stock of such developments a National Symposium on Materials and Processing -2012 (MAP-2012) was planned. The topics covered in the symposium are ceramics, glass/glass-ceramics and metals and materials. Papers relevant to INIS are indexed separately

  11. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

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

    International Nuclear Information System (INIS)

    The primary focus of this CRP was to perform detailed investigation of the high temperature industrial processes that are attainable through incorporation of an HTGR, and for their possible demonstration in the HTTR. The HTGR has the capability to achieve a core outlet temperature approaching 1,000 deg. C in a safe and effective manner. These attributes, coupled with the offer by JAERI to utilize the HTTR, resulted in the initiation of this CRP by the IAEA. High Temperature Engineering Test Reactor (HTTR) utilizes a 30 MW(th) HTGR comprised of 30 fuel columns of hexagonal pin-in-pin graphite block type fuel elements. The fuel consists of UO2 TRISO coated particles with an enrichment of ∼ 6% wt. Relative to the demonstration of high temperature heat applications, the HTTR will be capable of producing 10 MW(th) of heat at 950 deg. C. However, the thermal power for these applications has the potential to be increased up to 30 MW(th) in the future, which may be required for demonstration of gas turbine system components. The HTTR reached initial criticality in November 1998. Initial operational plans includes a series of rise to power tests followed by tests to demonstrate the safety and operational characteristics of the HTTR. In addition to completion of the HTTR demonstration tests, it was recommended that the R and D be performed within the HTTR project. JAERI is encouraged to publicize the results of the HTTR tests and 'lessons learned' from their experiences including potential capabilities of the HTGR for heat applications. The next priority application was determined to be the generation of electricity through the use of the gas turbine. Application of the Brayton Cycle utilizing high temperature helium from a modular HTGR was chosen for development because of its projected benefits as an economic and efficient means for the production of electricity. Evaluation of the remaining high temperature heat utilization applications chosen for investigation resulted

  13. Status of high temperature reactor development in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The Swedish AB Atomenergi held a HTR Information Seminar in Stockholm on January 11, 1978. At this seminar the status of the High Temperature Reactor development program in the Federal Republic of Germany was presented in a series of papers. This report containing the papers in english language makes the information also available to other parties outside Germany who are interested in the HTR development. The papers deal with the German HTR-program, the experience from construction and operation of HTR-plants, the HTR-technology, especially the nuclear system and its application for power production and process heat, the nuclear coal gasification and nuclear long distance energy transport, the HTR fuel cycles, the safety of HTR plants and nuclear process heat plants, an economic evaluation of the HTR as a power plant and/or source of process heat and the introduction strategy of this reactor system in the FRG. The review represents the status at the beginning of 1978. (orig.)

  14. PUMA - plutonium and minor actinides management in thermal high-temperature reactors

    International Nuclear Information System (INIS)

    The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6. Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled (HTR) reactor designs. The project runs from September 1, 2006 until August 31, 2009. The investigation on core physics aims at optimising the coated particle (CP) fuel and reactor characteristics, and assuring nuclear stability and safety of a Pu/Ma (minor actinides) HTR core. New CP designs will be explored in order to withstand very high burn-ups and obtain optimal adaptation for disposal after irradiation. In particular, helium production in Pu and MA-based fuel will be assessed and supported by experiments. Fuel irradiation performance codes, developed and used by several organisations, will permit convergence on optimized design criteria. The impact of the introduction of Pu/MA fuel on the fuel cycle and future energy mix will be assessed

  15. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

  16. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    International Nuclear Information System (INIS)

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value. - Highlights: • To validate the studies, a benchmark on core physics analysis was simulated using the SCALE6. • The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff. • The result presents good agreement with experimental value

  17. World development of nuclear power system and high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The major challenges of the nuclear power are economic competition, safety, proliferation resistance and waste storage. The safety of the nuclear power plants has been enhanced continuously in order to improve the public acceptance. The de-regulation of the electricity market in the United States and Europe encourages the competition between the different power generation technologies. The waste storage has been paid great attention. The ALWRs like ABWR, System 80+, EPR and AP-600 development from mid of 1980s are the major commercial products available in the market in the near future. When the time enter the 21 century, US DOE is planning to develop the Generation IV nuclear power system, in order to design on or more nuclear power systems that is market available before or in the year of 2020 and could be used to replace the current nuclear power plants. The modular pebble-bed High Temperature Gas-cooled Reactor (HTGR) is considered as a prefer candidate in the Generation IV nuclear power systems. The South Africa selected the modular HTGR as the reactor type to develop. The 10 MW test HTGR (HTR-10) designed and constructed in Tsinghua University Beijing is scheduled to reach critical in the year of 2000. Based on the HTR-10 project, China has established its preliminary capability of designing, constructing and manufacturing pebble-bed modular HTGRs

  18. Preliminary Core Analysis of High Temperature Engineering Test Reactor Using DeCART Code

    International Nuclear Information System (INIS)

    The 2-dimensional core analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with the ENDF-B/VII.0 library. From the analysis results, it is known that the DeCART code generally overestimates kinf with a moderator temperature variation. In addition, it can be seen that the DeCART code predicts less negative MTC than the McCARD code. However, the DeCART code gives a slightly more negative FTC value. From the depletion results, the error of the DeCART decreases over the burnup until 600 FPD. The DeCART code gives very similar trend within the error of 190 pcm, which is very small error when compared with other result

  19. Preliminary Core Analysis of High Temperature Engineering Test Reactor Using DeCART Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The 2-dimensional core analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with the ENDF-B/VII.0 library. From the analysis results, it is known that the DeCART code generally overestimates k{sub inf} with a moderator temperature variation. In addition, it can be seen that the DeCART code predicts less negative MTC than the McCARD code. However, the DeCART code gives a slightly more negative FTC value. From the depletion results, the error of the DeCART decreases over the burnup until 600 FPD. The DeCART code gives very similar trend within the error of 190 pcm, which is very small error when compared with other result.

  20. Safety aspect of high temperature nuclear reactor application for natural gas steam reforming

    International Nuclear Information System (INIS)

    An assessment of the safety aspect of high temperature nuclear reactor application for natural gas steam reforming has been carried out. The basic safety aspect associated with nuclear coupling to chemical process is to prevent the release of radioactive materials to the environment and or the chemical process. In utilizing nuclear heat for chemical process, intermediate heat exchanger (IHX) is used as an interface that separates nuclear and non nuclear zones. IHX is helium-helium heat exchanger in which the primary helium (905oC) coming out from the reactor, and transfer its heat to the secondary helium gas (890oC). To prevent possible release of radioactive materials from nuclear zone, balanced pressure is applied. The pressure of chemical process (4.5 MPa) is designed to be higher than the pressure of secondary helium (4.1 MPa) or primary helium (4 MPa). The design of balance pressure and the use of IHX cause some inferior condition of the nuclear heated reformer since the lower temperature (~800oC) reaches catalyst tube of reformer. This condition gives impact on lower thermal efficiency (~50%) compared to the fossil-fuelled plant (80-85%). Some modification in design and operation, such as: selecting the bayonet type of reformer equipped with orifice baffle, and enhancing heat utilization, can improve the lack of condition and are capable to increase the thermal efficiency of nuclear heated natural gas steam reformer to reach about 78%. (author)

  1. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    International Nuclear Information System (INIS)

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

  2. Combined constructal and exergy optimization of thermochemical reactors for high temperature heat storage

    International Nuclear Information System (INIS)

    Highlights: • Point to area flow optimization problem is solved for coupled heat and mass transfer. • Optimal solid/gas reactor geometry is found by minimizing the exergy destruction. • New defined properties permit easy pre-designing. • The method is applied to industrial high temperature thermal storage. - Abstract: High temperature heat storage is one of the key points for the development of solar power plants. Using reversible solid–gas chemical reactions is a promising solution to achieve high energy density and to reduce the storage volume. In order to achieve the high energy density, heat and mass transfer networks have to be optimized. In fact, such a reactive material presents antagonist behaviors for heat conductivity and gas permeability: increasing the reactive material density (i.e. the energy density) increases heat conductivity, but dramatically decreases permeability. An optimum has to be found. A method, combining constructal approach and exergy analysis is presented in this paper and applied to a solid/gas reactor, exchanging heat and matter (gas) with its surrounding. The gas is produced by the conversion of a solid S1 in a solid S2, implying a reaction heat. The method consists in evaluating the global entropy production of an elemental volume and minimizing it under two constraints: a given power density (kW/m3) and a given volume (i.e. given storage capacity), using Lagrange multipliers method. Then, a construction is done. The optimal shape and the number of elemental volumes constituting the reactor are searched. Taking into account heat and mass transfers, two networks emerge from the optimal construction: a heat conductive material network and a gas diffusers networks. The size of the conductive ‘fins’ and gas diffusers only depends on the properties of the reactive material (heat conductivity, permeability), the reactive gas (viscosity, pressure) and the heat of reaction. One important result is that global exergy

  3. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  4. Joining Characteristics of Intermediate Heat Exchanger Candidate Materials in Very High Temperature Reactor(VHTR)

    International Nuclear Information System (INIS)

    Worldwide studies have shown an increasing need for energy with the use of all energy sources, ranging from renewable sources through nuclear power, gas, to a limited extent oil and finally to the most prolific fossil fuel, coal. Although this increased need for generation capacity can met with different fuel sources, maybe the main fuel worldwide for next generation is hydrogen. The very high temperature reactor(VHTR) can produce hydrogen from only heat and water by using thermochemical iodine-sulfur(I-S) process or from heat, water, and natural gas by applying the steam reformer technology to core outlet temperatures greater than about 950 .deg. C. An intermediate heat exchanger(IHX) is the component in which the heat from the primary circuit helium is transferred to the secondary circuit helium(about 950 .deg. at 1000psi), thus keeping the secondary circuit free of radioactive contamination. The IHX will be located with a pressure vessel within the reactor containment that will be attached to the reactor pressure vessel by the cross-vessel. Therefore, an intermediate heat exchanger(IHX) especially is a key component in a VHTR. The Status of the IHX design will probably be a compact, counter-flow heat exchanger design consisting of metallic plate construction with small channels etched into each plate and assembled into a module. This heat exchanger design is refereed to as a 'printed circuit heat exchanger'. Printed circuit type heat exchanger are constructed from flat metal plates into which fluid flow channels are chemically milled. The milled plates are stacked and diffusion bonded together. In this study, the effects of the brazing temperature and homogenizing time for brazed specimens on the joint and base material microstructures, elemental distribution within the microstructures and the resulting joint tensile strength and micro hardness of Ni-based superalloy such as Haynes 230 were investigated

  5. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    International Nuclear Information System (INIS)

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

  6. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  7. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  8. Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

    International Nuclear Information System (INIS)

    Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

  9. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  10. Manufacturing and material properties of forgings for the reactor pressure vessel of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    For the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) which has been developed by the Japan Atomic Energy Research Institute (JAERI), 2.25Cr-1Mo steel is used for the first time in the world for a nuclear reactor pressure vessel. The RPV is 13.2 m in height and 5.5 m in internal diameter. Operation temperature is about 400 C and the internal pressure is 4 MPa. A material confirmation test has been carried out to demonstrate good applicability of forged low Si 2.25Cr-1Mo steel to the RPV of the HTTR. Recently, Japan Steel Works has succeeded in manufacturing large size ring forgings and a large size forged cover dome integrated with nozzles for the stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for the stand pipe. (orig.)

  11. Extension and application of the reactor dynamics code DYN3D for Block-type High Temperature Reactors

    International Nuclear Information System (INIS)

    The reactor code DYN3D was developed at the Helmholtz-Zentrum Dresden-Rossendorf to study steady state and transient behavior of Light Water Reactors. Concerning the neutronics part, the multigroup diffusion or SP3 transport equation based on nodal expansion methods is solved both for hexagonal and square fuel element geometry. To deal with Block-type High Temperature Reactor cores DYN3D was extended to a version DYN3D-HTR. A 3D heat conduction model was introduced to include 3D effects of heat transfer and heat conduction and the detailed structure of the fuel element. Homogenized neutronic cross sections were generated by applying a Monte Carlo approach with resolution of each individual TRISO fuel particle. Results of coupled steady state and transient calculations with 12 energy groups are presented. Transient case studies are control rod insertion, a change of the inlet coolant temperature and a change of the coolant gas mass flow rate. It is shown that DYN3D-HTR is an appropriate code system to simulate steady states and short time transients. Furthermore the necessity of the 3D heat conduction model is demonstrated

  12. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  13. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  14. Perspectives of modular high temperature gas-cooled reactor (MHTGR) on effluent management and siting

    International Nuclear Information System (INIS)

    The MHTGR is an advanced reactor concept being developed under a cooperative program involving the US Government, the utilities and the nuclear industry. The programs objective is the development of an environmentally safe, reliable, and economic nuclear power option for the USA and other nations of the world. HTGR design features, such as the ceramic fuel, helium coolant, and graphite moderator, are incorporated into the MHTGR reference plant design which incorporates four 350 MW(t) reactor modules. This papers objective is to describe those plant features, which minimize the environmental impact of MHTGR operation through efficient energy production, management of normal plant non-radioactive/radioactive effluents, and inherent characteristics and passive safety features which ensure benign plant site suitability source terms. 16 refs

  15. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  16. Friction and wear of ceramic pairs under high temperature conditions representative of advanced engine components

    International Nuclear Information System (INIS)

    Ball-on-disc friction and wear tests were performed with PSZ zirconia, Si3N4, and SiC ceramics and TiC cemented-carbide pairs under oscillating and linear sliding tests at 6500C in air and load conditions representative of advanced power systems. These tests showed high friction and wear of ceramic pairs at 6500C; improved performance was achieved coupling ceramics to TiC, and with TiC pairs. A review of practical lubrication systems for tribological engine components of high temperature materials showed that these exist and include solid lubrication, powder in gaseous carriers, and gas film support

  17. 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.

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

    International Nuclear Information System (INIS)

    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 9500C, 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 9500C 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

  19. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  20. Draft pre-application safety evaluation report for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    This draft safety evaluation report (SER) presents the preliminary results of a pre-application design review for the standard modular high-temperature gas-cooled reactor (MHTGR) (Project 672). The MHTGR conceptual design was submitted by the U.S. Department of Energy (DOE) in accordance with the U.S. Nuclear Regulatory Commission(NRC) 'Statement of Policy for the Regulation of Advanced Nuclear Power Plants' (51 FR 24643), which provides for early Commission review and interaction. The standard MHTGR consists of four identical reactor modules, each with a thermal output of 350 MWt, coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe. The reactors are helium cooled and graphite moderated and utilize ceramically coated particle-type nuclear fuel. The design includes passive reactor-shutdown and decay-heat-removal features. The staff and its contractors at the Oak Ridge National Laboratory and the Brookhaven National Laboratory have reviewed this design with emphasis on those unique provisions in the design that accomplish the key safety functions of reactor shutdown, decay-heat removal, and containment of radioactive material. This report presents the NRC staff's technical evaluation of those features in the MHTGR design important to safety, including their proposed research and testing needs. In addition this report presents the criteria proposed by the NRC staff to judge the acceptability of the MHTGR design and, where possible, includes statements on the potential of the MHTGR to meet these criteria. However, it should be recognized that final conclusions in all matters discussed in this report require approval by the Commission. Final determination on the acceptability of the MHTGR standard design is contingent on receipt and evaluation of additional information requested from DOE pertaining to the adequacy of the containment design and on the following: (1) satisfactory resolution of open safety issues identified

  1. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the

  2. Evaluation of the same heat Hastelloy XR as the material used for high-temperature components of the High-Temperature Engineering Test Reactor, 2

    International Nuclear Information System (INIS)

    A series of tension, Charpy impact and creep tests was carried out on two sorts of plate materials with 15 mm and 60 mm in thickness obtained from typical one of 30 heats of Hastelloy XR manufactured as the component material of the High-Temperature Engineering Test Reactor (HTTR). Creep test temperatures were 850, 900, 950 and 1000degC, and the maximum creep test time was 3371.4 h. The results obtained are as follows: (1) Both of plate materials tested exhibit acceptable tensile strength and tensile ductility as the structural material of the high-temperature components of the HTTR. (2) The plate material with 15 mm in thickness exhibits enough toughness, while toughness of the plate material with 60 mm in thickness is inferior to that of the plate material with 15 mm in thickness. (3) Both of plate materials tested possess the creep rupture strength beyond not only the expected minimum stress-to-rupture values, SR, but also the expected mean stress-to-rupture values of the material strength standards of Hastelloy XR. The materials also possess enough creep rupture ductility. (author)

  3. DELIGHT-6: one dimensional lattice burn-up code for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The code, DELIGHT-6, performs multi-group neutron spectrum calculation and provides few-group constans for succeeding core calculations. The main objective of the code is to serve as the lattice burn-up code for the core of a very high temperature gas-cooled reactor. The fuel rods of the reactor contain many coated fuel particles resulting double heterogeneous arrangement. The main calculational schema of DELIGHT-6 code is as follows; (1) Energy range for fast neutrons covers from 10 MeV to 2.38 eV and is divided into 61 fine groups. The thermal neutrons covers the rest of the energy range from 2.38 eV to 0 eV. Thermal spectrum is calculated by P1 or P0 approximation with 50 fine groups. (2) To treat resonance absorption, IR method is employed. (3) Zero and one dimensional models are available for the fuel lattice geometry and used for criticality and burn-up calculations. Collision probability method is adopted for the calculation of one dimensional model. (4) Shielding factor of burnable poison is calculated by collision probability method. (5) Other functions of the code are; 1. Spatial shielding factor calculation of 240Pu, 2. Calculation of neutron streaming effect caused by a gap or a hole in the fuel lattice, 3. Calculation of neutron flux distribution in the fuel lattice by diffusion theory, 4. Calculation of Xe and Sm absorption cross sections with burn-up. (6) Cross section library in both fast and thermal energy range is compiled from ENDF/B-4 except burn-up data of Xm, Sm and pseudo FPs which are supplied by ENDF/B-3. (7) The code provides the macroscopic group constants of fuel lattice with burn-up in CITATION input format. (jin)

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

    International Nuclear Information System (INIS)

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

  5. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 1000C to 7600C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. The insulation retained an acceptable degree of resiliency. However, some fiber damage was observed within both the high and low temperature insulation blankets. A thermal analysis was conducted to correlate the hot duct heat transfer results with those obtained from the analytical techniques used for the HTGR design using a computer thermal model representative of the duct and test setup. The thermal performance of the insulation, the temperature gradient through the structural components, the heating load to the cooling system and the permeation flow effect on heat transfer were verified. Exellent correlation between the experimental data and the analytical techniques were obtained

  6. Synthetic fuel production using Texas lignite and a very high temperature reactor for process heat

    International Nuclear Information System (INIS)

    Two approaches for synthetic fuel production from coal are studied using Texas lignite as the feedstock. First, the gasification and liquefaction of coal are accomplished using Lurgi gasifiers and Fischer-Tropsch synthesis. A 50 000 barrel/day facility, consuming 13.7 million tonne/yr (15.1 million ton/yr) of lignite, is considered. Second, a nuclear-assisted coal conversion approach is studied using a very high temperature gas-cooled reactor with a modified Lurgi gasifier and Fischer-Tropsch synthesis. The nuclear-assisted approach resulted in a 35% reduction in coal consumption. In addition, process steam consumption was reduced by one-half and the oxygen plants were eliminated in the nuclear assisted process. Both approaches resulted in a synthetic oil price higher than the March 1980 imported price of $29.65 per barrel: $36.15 for the lignite-only process and $35.16 for the nuclear-assisted process. No tax advantage was assumed for either process and the utility financing method was used for both economic calculations

  7. Corrosion of high temperature alloys in the coolant helium of a gas cooled reactor

    International Nuclear Information System (INIS)

    The corrosion of structural alloys in gas cooled reactor environment appears to be a critical issue. The coolant helium proved to contain impurities mainly H2, H2O, CO, and CH4 in the microbar range that interact with metallic materials at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the gas chemistry, the alloy composition and the temperature. These structural transformations can notably influence the component mechanical properties. A short review of the literature on the topic is first given. Corrosion tests with high chromium alloys and a Mo-based alloy were carried out at 750 C in a purposely-designed facility under simulated GCR helium. The first, rather short term, results showed that the Mo-based alloy was inert while the others alloys oxidised during at least 900 hours. The alloy with the higher Al and Ti contents exhibited poor oxidation resistance impeding its use as structural material without further investigations. (orig.)

  8. Techno-economic analysis of seawater desalination using high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Our world, including China (especially in big cities and foreland), is facing the increased global shortage of potable water and pollution of water. It is ideal to promote seawater desalination to satisfy the potable water demand in these areas. Among the various processes, MED, RO and VC have proven well developed and promising. Due to the inherent safety and its vapor produced with high parameters and features of small size and modular design, HTGR (High Temperature Gas-cooled Reactor) of 2x200MW is chosen as the energy source for the desalination in dual production of clean water and power. This paper discusses the techno-economic feasibility of different seawater desalting systems using 2x200MW HTGR in the areas mentioned above, that is, ST-MED (Steam Turbine Cycle), RO, MED/TVC, RO/MED and GT-MED (Gas Turbine Cycle). The exergy concept is used in calculating availability to get cost of energy in desalination, and power credit method is used in economic assessment of different systems to get reasonable evaluating, while economic-life levelized cost method is adopted for calculating electricity cost of referred HTGR plant. In addition, sensitivity analysis on ST-MED economy is also presented. (author)

  9. Properties influencing high-temperature gas-cooled reactor coated fuel particle performance

    International Nuclear Information System (INIS)

    Properties affecting the irradiation performance of outer pyrolytic carbon (PyC) layers on Triso- and Biso-coated fuel particles were studied. Irradiation temperatures were 1000 to 15000C (1273 to 1773 K). Fast-neutron fluences reached 12.4 x 1025 n/m2 (E greater than 29 fJ)/sub HTGR/, which is 55 percent beyond the large high-temperature gas-cooled reactor peak design exposure of 8.0 x 1025 n/m2. Coatings with densities between 1.85 and 1.95 Mg/m3 and mean optical anisotropy values of equal to or less than 1.03 (BAF0 units) exhibited the best irradiation performance on Triso particles. For Biso particles, it is necessary to deposit the outer layer at coating rates between 3 and 8 μm/min and with densities equal to or greater than 1.84 Mg/m3 to produce coatings impermeable to fission gases after irradiation. Data from fuel rod tests show that it is important to limit the degree of surface-connected porosity of the outer PyC layer and the amount of binder phase in the matrix to prevent coating failures resulting from coating-matrix interactions

  10. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    International Nuclear Information System (INIS)

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

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

    International Nuclear Information System (INIS)

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

  12. Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

    Institute of Scientific and Technical Information of China (English)

    CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

    2006-01-01

    Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor (HTR). The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUT-DGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP'94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239Pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

  13. Development of fuel failure detection system for a High Temperature Gas Cooled Reactor (V)

    International Nuclear Information System (INIS)

    This paper reports on a fuel failure detection (FFD) system using a wire-precipitator developed for a High Temperature Gas cooled reactor (HTGR). On actual application of the FFD, it is important to inquire the response characteristics of the precipitator and the behavior of noble-gas-FPs released from coated particle fuel compacts. The dependence of the precipitator counting rate on purge-gas flow-rate was measured. A response function of the precipitator including the dilution effect of the purge-gas was fabricated. Adsorption characteristics of a charcoal-filter for noble-gas-FPs was measured. Under the low flow-rate, noble-gas-FPs are adsorbed by the charcoal-filter and Xenon are adsorbed easily. A preliminary experiment for FFD system using this adsorption effect was performed. Moreover, a FFD response function for noble-gas-FPs circulated in a primary coolant system was developed to estimate the released noble-gas-FPs in the primary coolant system. The validity of this function was confirmed by experiments using the Helium gas loop OGL-1

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

    International Nuclear Information System (INIS)

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

  15. Potential applications of helium-cooled high-temperature reactors to process heat use

    International Nuclear Information System (INIS)

    High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes

  16. Benchmark analysis of high temperature engineering test reactor core using McCARD code

    International Nuclear Information System (INIS)

    A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% Δk/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % Δk/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)

  17. Studies of bearings to be used in high-temperature reactor plants

    International Nuclear Information System (INIS)

    Several components of high-temperature reactor units have roller bearings in order to carry out their motions without much friction. The use of such bearings poses friction and wear problems which cannot be mastered by commercial roller bearing technology. Possible improvements of coating, cage design and bearing materials as well as of their parameters were registered and studied. The service life of dry lubricated bearings was considerably improved. With a radial or axial load on the bearing of ≥ 10% of the static load, ≅ 20x106 rolling motions/actuations can be performed. The connections between surface compression and wear were determined, and optimum conditions for the transfer of lubricants from the cage onto the bearing race were worked out. Coating of corrosion-resistant roller bearing steels with the HRB-M0S2 running-in coating could be proved. New cage designs and materials were tested with positive results. Alternative coatings (thin chromium layer) and lubricants (SLC, MCR, PCR) were tested. (orig./HP) With 56 figs

  18. Remarks on the thermochemical production of hydrogen from water using heat from the high temperature reactor

    International Nuclear Information System (INIS)

    In this report, some aspects of the production of hydrogen from water using heat from the High Temperature Reactor has been studied. These aspects are: the theoretical potential for economic competitivness, the application of hydrogen in the Heat Market, the size of the market potential in the Federal Republic of Germany and the extent of research and development work. In addition another novel proposal for a thermochemical cycle has been studied. For the description of the theoretical potential for economic competitivness, a definition of the 'coupling', has been introduced, which is thermodynamicaly developed; the thermochemical cycle is compared with the thermochemical cycle. Using the coupling, it becomes possible to describe a relation between thermodynamical parameters and the ecomomical basic data of capital costs. Reasons are given from the theoretical point of view for the application of hydrogen as an energy carrier of high exergetic value in the heat market. The discussion of energy problems as 'questions of global survival' leads here to a proposal for the introduction of the term 'extropy'. The market potential in the Federal Republic of Germany is estimated. A further novel proposal for a thermochemical cycle is the 'hydrocarbon-hybrid-process'. The extent of research and development work is explained. (orig.)

  19. Depressurization accidents in a medium-sized high-temperature gas reactor

    International Nuclear Information System (INIS)

    The amount of fission product release during a core heatup accident in a medium-sized high-temperature gas reactor depends on the size of the inadvertent opening in the primary circuit; this dependence is assessed. The opening triggers a depressurization event that is assumed to be coupled with the failure of the forced circulation in both decay-heat removal systems. The scenario investigated is a beyond-design-base accident. The DSNP modular simulation code is used. This paper reports that a two-dimensional model is developed to simulate the HTR-500 design. The study shows that the depressurization process does not contribute significantly to the sweeping out (from the primary circuit) of fission products released from the fuel during the core heatup. There is also no significant variation in the results when the opening size is >33 cm2, and only a slight sensitivity is found when the rupture size is between 3.3 and 33 cm2. The fission product release decreases considerably in the range from 1 to 3.3 cm2. The small-sized rupture is of major significance, as the failure of the relief valves to reclose increases the frequency of the event

  20. Application of the High Temperature Gas Cooled Reactor to oil shale recovery

    International Nuclear Information System (INIS)

    Current oil shale recovery processes combust some portion of the products to provide energy for the recovery process. In an attempt to maximize the petroleum products produced during recovery, the potentials for substituting nuclear process heat for energy generated by combustion of petroleum were evaluated. Twelve oil shale recovery processes were reviewed and their potentials for application of nuclear process heat assessed. The High Temperature Gas Cooled Reactor-Reformer/Thermochemical Pipeline (HTGR-R/TCP) was selected for interfacing process heat technology with selected oil shale recovery processes. Utilization of these coupling concepts increases the shale oil product output of a conventional recovery facility from 6 to 30 percent with the same raw shale feed rate. An additional benefit of the HTGR-R/TCP system was up to an 80 percent decrease in emission levels. A detailed coupling design for a typical counter gravity feed indirect heated retorting and upgrading process were described. Economic comparisons prepared by Bechtel Group Incorporated for both the conventional and HTGR-R/TCP recovery facility were summarized