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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. 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.

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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.)

  10. 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

  11. 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.)

  12. 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

  13. 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

  14. 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)

  15. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-01

    This report 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 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.

  16. 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

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. Advanced characterizations of austenitic oxide dispersion-strengthened (ODS) steels for high-temperature reactor applications

    Science.gov (United States)

    Miao, Yinbin

    Future advanced nuclear systems involve higher operation temperatures, intenser neutron flux, and more aggressive coolants, calling for structural materials with excellent performances in multiple aspects. Embedded with densely and dispersedly distributed oxide nanoparticles that are capable of not only pinning dislocations but also trapping radiation-induced defects, oxide dispersion-strengthened (ODS) steels provide excellence in mechanical strength, creep resistance, and radiation tolerance. In order to develop ODS steels with qualifications required by advanced nuclear applications, it is important to understand the fundamental mechanisms of the enhancement of ODS steels in mechanical properties. In this dissertation, a series of austenitic ODS stainless steels were investigated by coordinated state-of-the-art techniques. A series of different precipitate phases, including multiple Y-Ti-O, Y-Al-O, and Y-Ti-Hf-O complex oxides, were observed to form during mechanical alloying. Small precipitates are likely to have coherent or cubic-on-cubic orientation relationships with the matrix, allowing the dislocation to shear through. The Orowan looping mechanism is the dominant particle-dislocation interaction mode as the temperature is low, whereas the shearing mechanism and the Hirsch mechanism are also observed. Interactions between the particles and the dislocations result in the load-partitioning phenomenon. Smaller particles were found to have the stronger loading-partitioning effect. More importantly, the load-partitioning of large size particles are marginal at elevated temperatures, while the small size particles remain sustaining higher load, explaining the excellent high temperature mechanical performance of ODS steels.

  2. 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

  3. 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.

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. Fusion reactors-high temperature electrolysis (HTE)

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A. (ed.)

    1978-01-01

    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 1800/sup 0/C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400/sup 0/C 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 H/sub 2/ 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 10/sup 6/ 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.

  10. 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

  11. 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.

  12. 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.

  13. 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

  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. 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.

  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. 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.)

  18. 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

  19. Ultra high temperature particle bed reactor design

    Science.gov (United States)

    Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

    1990-01-01

    A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

  20. 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

  1. 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.)

  2. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  3. 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)

  4. 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.

  5. 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

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. High-temperature reactor developments in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    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.).

  16. 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.)

  17. 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)

  18. 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.

  19. 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...

  20. 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 ...

  1. 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)

  2. 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.

  3. 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

  4. 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.

  5. 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)

  6. 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).

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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

  13. 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.)

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    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. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  5. 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)

  6. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T. (University of Southern California, Los Angeles, CA (United States). Dept. of Chemical Engineering); Liu, P.K.T. (Aluminum Co. of America, Pittsburgh, PA (United States)); Webster, I.A. (Unocal Corp., Los Angeles, CA (United States))

    1992-01-01

    Membrane reactors are today finding extensive applications for gas and vapor phase catalytic reactions (see discussion in the introduction and recent reviews by Armor [92], Hsieh [93] and Tsotsis et al. [941]). There have not been any published reports, however, of their use in high pressure and temperature liquid-phase applications. The idea to apply membrane reactor technology to coal liquid upgrading has resulted from a series of experimental investigations by our group of petroleum and coal asphaltene transport through model membranes. Coal liquids contain polycyclic aromatic compounds, which not only present potential difficulties in upgrading, storage and coprocessing, but are also bioactive. Direct coal liquefaction is perceived today as a two-stage process, which involves a first stage of thermal (or catalytic) dissolution of coal, followed by a second stage, in which the resulting products of the first stage are catalytically upgraded. Even in the presence of hydrogen, the oil products of the second stage are thought to equilibrate with the heavier (asphaltenic and preasphaltenic) components found in the feedstream. The possibility exists for this smaller molecular fraction to recondense with the unreacted heavy components and form even heavier undesirable components like char and coke. One way to diminish these regressive reactions is to selectively remove these smaller molecular weight fractions once they are formed and prior to recondensation. This can, at least in principle, be accomplished through the use of high temperature membrane reactors, using ceramic membranes which are permselective for the desired products of the coal liquid upgrading process. An additional incentive to do so is in order to eliminate the further hydrogenation and hydrocracking of liquid products to undesirable light gases.

  7. 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)

  8. 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)

  9. 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.)

  10. 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.

  11. 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)

  12. 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

  13. 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

  14. 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

  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. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-03

    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

  18. 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)

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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)

  6. 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)

  7. High-temperature membrane reactors: potential and problems

    NARCIS (Netherlands)

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

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours (Saracco et al., 1994), describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. S

  8. 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

  9. Current hurdles to the success of high temperature membrane reactors

    NARCIS (Netherlands)

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

    1994-01-01

    High-temperature catalytic processs performed using inorganic membranes have been in recent years a fast growing area of research, which seems to have not yet reached its peak. Chemical engineers, catalysts and materials scientists have addressed this topic from different viewpoint in a common effor

  10. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  18. Tritium Formation and Mitigation in High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  19. 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

  20. 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)

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

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

    Science.gov (United States)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    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.

  11. 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.

  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); 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.

  13. 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.

  14. 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.

  15. 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

  16. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  17. 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

  18. 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.

  19. 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)

  20. 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.)

  1. 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.

  2. 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)

  3. 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.

  4. Advanced energy analysis of high temperature fuel cell systems

    NARCIS (Netherlands)

    De Groot, A.

    2004-01-01

    In this thesis the performance of high temperature fuel cell systems is studied using a new method of exergy analysis. The thesis consists of three parts: ⢠In the first part a new analysis method is developed, which not only considers the total exergy losses in a unit operation, but which distingu

  5. 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

  6. Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

    Science.gov (United States)

    Hall, Michael L.; Doster, Joseph M.

    1986-01-01

    Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

  7. 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...

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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)

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

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

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

  20. 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.

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. Design of fuzzy PID controller for high temperature pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Badgujar, Kushal D.; Satpute, Satchidanand R.; Revankara, Shripad T.; Lee, John C.; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of)

    2012-10-15

    Control system is most important characteristic to be considered to control spontaneous fission reaction in the design of the nuclear reactor. Recently fuzzy based control systems have been designed and applied as control system for nuclear plants. This article emphasize on controlling the power of the high temperature pebble bed reactor (HTPBR) with the design of Fuzzy proportional integral derivative (PID) controller. A simplified reactor model with point kinetics equation and reactor heat balance equation is used. The reactivity feedback arising from power coefficient of reactivity and Xenon poisoning is also considered. The reactor is operated at various power levels by using fuzzy PID controller. The fuzzy logic eliminates the necessity of the tuning the gains of PID controller each time by extending the finite sets of the PID controller gains.

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Losa, Evžen, E-mail: evzen.losa@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Souček, Václav; Kohout, Petr [AZIN CZ, s.r.o., Hanusova 3, 140 00 Praha 4 (Czech Republic)

    2014-05-01

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

  16. 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

  17. 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.

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. 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.)

  3. 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.

  4. 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

  5. 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...

  6. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Science.gov (United States)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  7. 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.

  8. 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.)

  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. 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)

  11. 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)

  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. 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

  14. 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

  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. 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)

  17. 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)

  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. 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)

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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.

  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. 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.

  10. 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)

  11. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  12. 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.

  13. The use of a very high temperature nuclear reactor in the manufacture of synthetic fuels

    Science.gov (United States)

    Farbman, G. H.; Brecher, L. E.

    1976-01-01

    The three parts of a program directed toward creating a cost-effective nuclear hydrogen production system are described. The discussion covers the development of a very high temperature nuclear reactor (VHTR) as a nuclear heat and power source capable of producing the high temperature needed for hydrogen production and other processes; the development of a hydrogen generation process based on water decomposition, which can utilize the outputs of the VHTR and be integrated with many different ultimate hydrogen consuming processes; and the evaluation of the process applications of the nuclear hydrogen systems to assess the merits and potential payoffs. It is shown that the use of VHTR for the manufacture of synthetic fuels appears to have a very high probability of making a positive contribution to meeting the nation's energy needs in the future.

  14. 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

  15. 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.)

  16. Legal obstacles to the construction of high temperature reactors for heat generation on the example of Polish regulations

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, Jan; Koszuk, Lukasz; Klisinska, Malgorzata; Andrzejewski, Krzysztof [National Centre for Nuclear Research, Otwock (Poland). Nuclear Energy Div.

    2016-07-15

    High temperature gas-cooled reactors (HTR) can produce high temperature process heat. This extends their potential range of application. In recent decades, the practice of licensing reactors of HTR technology, caused by the need of streamlining of this process, has placed the main emphasis on the adaptation of existing regulations by formulating appropriate guidelines. Currently, in-depth analyses of usefulness and shortcomings of the existing regulations are required. The article analyzes the possibility of licensing of high-temperature reactors with cogeneration based on the Polish Atomic Law and the three resolutions of the Council of Ministers.

  17. 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.

  18. 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

  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. 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.

  1. 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.)

  2. 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.

  3. 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.

  4. 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

  5. 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)

  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. 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

  8. 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.

  9. 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)

  10. 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

  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. 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

  13. 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.

  14. 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)

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  20. High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

    Science.gov (United States)

    Roman, W. C.

    1979-01-01

    An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

  1. 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)

  2. 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)

  3. 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.

  4. 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)

  5. 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

  6. 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.

  7. 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)

  8. 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.

  9. 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

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of); Lee, Young-woo; Cho, Moon-sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    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

  12. 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.)

  13. 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.

  14. 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)

  15. Application of a moving bed biofilm reactor for tertiary ammonia treatment in high temperature industrial wastewater.

    Science.gov (United States)

    Shore, Jennifer L; M'Coy, William S; Gunsch, Claudia K; Deshusses, Marc A

    2012-05-01

    This study examines the use of a moving bed biofilm reactor (MBBR) as a tertiary treatment step for ammonia removal in high temperature (35-45°C) effluents, and quantifies different phenotypes of ammonia and nitrite oxidizing bacteria responsible for nitrification at elevated temperatures. Bench scale reactors operating at 35 and 40°C were able to successfully remove greater than 90% of the influent ammonia (up to 19 mg L(-1) NH(3)-N) in both the synthetic and industrial wastewater. No biotreatment was observed at 45°C, although effective nitrification was rapidly recovered when the temperature was lowered to 30°C. Using qPCR, Nitrosomonas oligotropha was found to be the dominant ammonia oxidizing bacterium in the biofilm for the first phases of reactor operation. In the later phases, Nitrosomonas nitrosa was observed and its increased presence may have been responsible for improved ammonia treatment efficiency. Accumulation of nitrite in some instances appeared to correlate with temporary low presence of Nitrospira spp.

  16. 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

  17. 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

  18. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  19. 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

  20. 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

  1. Novel high-temperature reactors for in situ studies of three-way catalysts using turbo-XAS.

    Science.gov (United States)

    Guilera, Gemma; Gorges, Bernard; Pascarelli, Sakura; Vitoux, Hugo; Newton, Mark A; Prestipino, Carmelo; Nagai, Yasutaka; Hara, Naoyuki

    2009-09-01

    Two novel high-temperature reactors for in situ X-ray absorption spectroscopy (XAS) measurements in fluorescence are presented, each of them being optimized for a particular purpose. The powerful combination of these reactors with the turbo-XAS technique used in a dispersive-XAS beamline permits the study of commercial three-way catalysts under realistic gas composition and temporal conditions.

  2. 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

  3. 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)

  4. 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)

  5. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    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.

  6. 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

  7. Simultaneous approach for simulation of a high-temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Yang CHEN; Jiang-hong YOU; Zhi-jiang SHAO; Ke-xin WANG; Ji-xin QIAN

    2011-01-01

    The simulation of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations (DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail; both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

  8. 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)

  9. 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.

  10. 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).

  11. 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)

  12. 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

  13. 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

  14. 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.

  15. 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.)

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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)

  1. 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)

  2. 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)

  3. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    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.

  4. 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

  5. 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...

  6. 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...

  7. 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

  8. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  3. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  11. 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

  12. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. Thorium–based fuel cycles : saving uranium in a 200 MWth pebble bed high temperature reactor / S.K. Gintner

    OpenAIRE

    Gintner, Stephan Konrad

    2010-01-01

    The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In additio...

  20. 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

  1. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Dunfu

    2015-07-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  2. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  3. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

  4. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    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 project. Because the NRC 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). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: (1) The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. (2) The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. (3) Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW

  5. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    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 project. Because the NRC 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). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  6. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Battaglia, Francine

    2008-11-29

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  7. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  8. An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Highlights: • An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor was developed. • The optimized process and verification experiment using the HTR-10 were presented in detail. • A large quantity of high-dose tritium-containing waste water was successfully collected in commissioning experiment of the improved HTR-10. • The optimized process was proved to be reliable to avoid the large emission of radioactive waste water to the environment. - Abstract: An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor (HTGR) was developed and experimentally verified using the 10 MW High-Temperature Gas-cooled Reactor-test module (HTR-10). Compared with the previous process, an auxiliary molecular sieve bed was added in helium purification regeneration system and new operation process was proposed to collect tritium-containing waste water. In this paper, the optimized process and verification experiment were presented in detail. In commissioning experiment of the improved HTR-10, a large quantity of high-dose tritium-containing waste water was successfully collected in the water separator of helium purification regeneration system, with the specific activity being 6.1 × 109 Bq/L. The verification experiment confirms that the optimized process is effective and reliable for the demonstration plant design of High Temperature Gas-cooled Reactor-Pebble bed module (HTR-PM) to avoid the large emission of detrimentally radioactive waste water to the environment

  9. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  10. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  11. A kinetic study of gaseous potassium capture by coal minerals in a high temperature fixed-bed reactor

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Jensen, Peter Arendt; Jensen, Anker Degn

    2008-01-01

    The reactions between gaseous potassium chloride and coal minerals were investigated in a lab-scale high temperature fixed-bed reactor using single sorbent pellets. The applied coal minerals included kaolin, mullite, silica, alumina, bituminous coal ash, and lignite coal ash that were formed...

  12. ICONE-4: Proceedings. Volume 2: Advanced reactors

    International Nuclear Information System (INIS)

    The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

  13. Active chemistry control for coolant helium applying high-temperature gas-cooled reactors - HTR2008-58096

    International Nuclear Information System (INIS)

    Lifetime extension of high-temperature equipment such as the intermediate heat exchanger of high-temperature gas-cooled reactors (HTGRs) is important from the economical point of view. Since the replacing cost will cause the increasing of the running cost, it is important to reduce replacing times of the high-cost primary equipment during assumed reactor lifetime. In the past, helium chemistry has been controlled by the passive chemistry control technology in which chemical impurity in the coolant helium is removed as low concentration as possible, as does Japan's HTTR. Although the lifetime of high- temperature equipment almost depends upon the chemistry conditions in the coolant helium, it is necessary to establish an active chemistry control technology to maintain adequate chemical conditions. In this study, carbon deposition which could occur at the surface of the heat transfer tubes of the intermediate heat exchanger and decarburization of the high-temperature material of Hastelloy XR used at the heat transfer tubes were evaluated by referring the actual chemistry data obtained by the HTTR. The chemical equilibrium study contributed to clarify the algorism of the chemistry behaviours to be controlled. The created algorism is planned to be added to the instrumentation system of the helium purification systems. In addition, the chemical composition to be maintained during the reactor operation was proposed by evaluating not only core graphite oxidation but also carbon deposition and decarburization. It was identified when the chemical composition could not keep adequately, injection of 10 ppm carbon monoxide could effectively control the chemical composition to the designated stable area where the high-temperature materials could keep their structural integrity beyond the assumed duration. The proposed active chemistry control technology is expected to contribute economically to the purification systems of the future very high-temperature reactors. (authors)

  14. Analysis of Possible Application of High-Temperature Nuclear Reactors to Contemporary Large-Output Steam Power Plants on Ships

    Directory of Open Access Journals (Sweden)

    Kowalczyk T.

    2016-04-01

    Full Text Available This paper is aimed at analysis of possible application of helium to cooling high-temperature nuclear reactor to be used for generating steam in contemporary ship steam-turbine power plants of a large output with taking into account in particular variable operational parameters. In the first part of the paper types of contemporary ship power plants are presented. Features of today applied PWR reactors and proposed HTR reactors are discussed. Next, issues of load variability of the ship nuclear power plants, features of the proposed thermal cycles and results of their thermodynamic calculations in variable operational conditions, are presented.

  15. Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies.

    Science.gov (United States)

    Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

    2013-03-01

    Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ∼10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified. PMID:23556828

  16. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  17. Suitability of Cadmium Tantalate and Indium Tantalate as Control Materials for High-Temperature Reactors

    International Nuclear Information System (INIS)

    Control materials for practical use in high-temperature reactors should, independently of the requirements of the individual case, have the following properties: (a ) high absorption cross-section for neutrons in a wide range of energies; (b ) high absorption capacity for neutrons; (c ) small sensitivity for radiation damage; (d) good thermal resistance; (e ) low reactivity with the environment; and ( f ) low costs and good availability. With these points and the avoidance of the disadvantages of n, α -reactions taken into consideration, attention should be paid chiefly to the elements cadmium, tungsten, indium and tantalum. It is important to combine a good thermal absorber with an epithermál absorber so that the resulting material is stable at elevated temperatures (≳ 700°C). For this purpose the double-oxides CdWO4, Cd 2Ta2O7 and CdIn2O2 are suitable. Among these, cadmium tantalate has the highest thermal resistance. Another double-oxide which in combination with cadmium tantalate possesses an advantageous absorption spectrum for neutrons is indium tantalate. It has also good thermal resistance. Because ceramic absorber materials often have to be shaped by plastic deformation, they usually are used as cermets. Therefore, they must be compatible with metals. Cadmium tantalate is compatible with silver and copper and up to 700°C with nickel; indium, tantalate is completely compatible with silver, copper and nickel and up to 700°C with molybdenum also and to some degree with iron. These results are in agreement with thermodynamical calculations. For an estimation of the behaviour of the absorber materials under reactor conditions the daughter products originating from neutron absorption have to be considered. While Cd113 transforms into the stable Cd114, tantalum transmutes into tungsten and indium into tin. Both daughter products can bind more oxygen in their most stable valency states than the parent elements can. Therefore, the reduction of Cd++to metal

  18. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    S. Yamoah

    2012-06-01

    Full Text Available The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as thermal convection and conduction. In this study, a model with the capability to simulate fluid flow and heat transfer in the pebble bed modular reactor core has been developed. The developed model was implemented on a personal computer using FORTRAN 95 programming language. Several important fluid flow and heat transfer parameters have been examined: including the pressure drop over the reactor core, the heat transfer coefficient, the Nusselt number and the effective thermal conductivity of the fuel pebbles. Results obtained from the simulation experiments show a uniform pressure in the radial direction for a core to fuel element diameter (D/d ratio>20 and the heat transfer coefficient increases with increasing temperature and coolant mass flow rate. The model can adequately account for the flow and heat transfer phenomenon and the loss of pressure through friction in the pebble bed type high temperature nuclear reactor.

  19. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  20. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2012-07-01

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. 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 first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

  1. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  2. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  3. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

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

    International Nuclear Information System (INIS)

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

  5. Advance in highly efficient hydrogen production by high temperature steam electrolysis

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    High Temperature Steam Electrolysis (HTSE) through a solid oxide electrolytic cell (SOEC) has been receiving increasing research and development attention worldwide because of its high conversion efficiency (about 45%-59%) and its potential usage for large-scale production of hydrogen. The mechanism, composition, structure, and developing challenges of SOEC are summarized. Current situation, key materials, and core technologies of SOEC (solid oxide electrolytic cell) in HTSE are re- viewed, and the prospect of HTSE future application in advanced energy fields is proposed. In addition, the recent research achievements and study progress of HTSE in Tsinghua University are also intro- duced and presented.

  6. Advance in highly efficient hydrogen production by high temperature steam electrolysis

    Institute of Scientific and Technical Information of China (English)

    YU Bo; ZHANG WenQiang; CHEN Jing; XU JingMing; WANG ShaoRong

    2008-01-01

    High Temperature Steam Electrolysis (HTSE) through a solid oxide electrolytic cell (SOEC) has been receiving increasing research and development attention worldwide because of its high conversion efficiency (about 45%-59%) and its potential usage for large-scale production of hydrogen. The mechanism, composition, structure, and developing challenges of SOEC are summarized. Current situation, key materials, and core technologies of SOEC (solid oxide electrolytic cell) in HTSE are re-viewed, and the prospect of HTSE future application in advanced energy fields is proposed. In addition, the recent research achievements and study progress of HTSE in Tsinghua University are also intro-duced and presented.

  7. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  8. High Temperature Materials Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The High Temperature Materials Lab provides the Navy and industry with affordable high temperature materials for advanced propulsion systems. Asset List: Arc Melter...

  9. The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H.; Schuerrer, F.; Ninaus, W.; Oswald, K.; Rabitsch, H.; Kreiner, H. (Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik); Neef, R.D. (Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung)

    1984-12-15

    To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  11. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  12. Rigid bonded glass ceramic seals for high temperature membrane reactors and solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Paulsen, Ove

    2009-05-15

    Solid Oxide Fuel cells (SOFC) and dense gas separation membranes based on mixed ionic and electronic conductors have gained increased interest the resent years due the search for new technologies for clean energy generation. These technologies can be utilized to produce electricity from fossil fuel with low CO{sub 2} emission compared to conventional gas or coal based energy plants. One crucial challenge with high temperature membrane reactors and SOFCs is the sealing of the active membranes/electrolytes to prevent leakage of air to fuel side or vice versa. Due to the high operating temperatures of typical 800-1000 degrees Celsius the selection of reliable sealing materials is limited. The seals have to remain gas tight during the life time of the reactor/SOFC, they need to be chemical compatible with the sealed materials and stable in reducing and oxidizing atmospheres containing water vapour and CO{sub 2}, and finally they should be cheap, readily available and easy to process. The main purpose of the present work was to evaluate rigid bonded glass ceramic seals for dense oxygen ion and proton conducting membranes and electrolytes for SOFCs and high temperature (HT) membrane reactors. First, a review of sealing technologies has been carried out with emphasis on SOFC and ceramic membranes technologies applicable for zero emission power plants. Regarding sealing, the best and cheapest materials at the present time are based on silicate glass and glass ceramics. In the present work aluminate glass without silica is introduced as a new class of seals expanding the material selection for HT membrane sealing technologies. The main reason for studying silica free systems is that silica is known to be unstable in humid atmospheres and/or reducing conditions at elevated temperatures. Two glass systems have been evaluated. The first was based on aluminate glasses in the system RO-CaO-Al{sub 2}O{sub 3} (R=Mg, Ba, Sr) with special focus on the CaO-MgO-Al{sub 2}O{sub 3

  13. Advanced Materials for High Temperature, High Performance, Wide Bandgap Power Modules

    Science.gov (United States)

    O'Neal, Chad B.; McGee, Brad; McPherson, Brice; Stabach, Jennifer; Lollar, Richard; Liederbach, Ross; Passmore, Brandon

    2016-01-01

    Advanced packaging materials must be utilized to take full advantage of the benefits of the superior electrical and thermal properties of wide bandgap power devices in the development of next generation power electronics systems. In this manuscript, the use of advanced materials for key packaging processes and components in multi-chip power modules will be discussed. For example, to date, there has been significant development in silver sintering paste as a high temperature die attach material replacement for conventional solder-based attach due to the improved thermal and mechanical characteristics as well as lower processing temperatures. In order to evaluate the bond quality and performance of this material, shear strength, thermal characteristics, and void quality for a number of silver sintering paste materials were analyzed as a die attach alternative to solder. In addition, as high voltage wide bandgap devices shift from engineering samples to commercial components, passivation materials become key in preventing premature breakdown in power modules. High temperature, high dielectric strength potting materials were investigated to be used to encapsulate and passivate components internal to a power module. The breakdown voltage up to 30 kV and corresponding leakage current for these materials as a function of temperature is also presented. Lastly, high temperature plastic housing materials are important for not only discrete devices but also for power modules. As the operational temperature of the device and/or ambient temperature increases, the mechanical strength and dielectric properties are dramatically reduced. Therefore, the electrical characteristics such as breakdown voltage and leakage current as a function of temperature for housing materials are presented.

  14. Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Philip Casey Durst; Mark Schanfein

    2012-08-01

    The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information

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

    International Nuclear Information System (INIS)

    Ten years ago, the European High Temperature Reactor (HTR) Technology Network (HTR-TN) launched a programme for developing HTR Technology, which expanded so far through 4 successive Euratom Framework Programmes. Many projects have been performed - in particular the RAPHAEL project in the 6th Euratom Framework Programme and presently ARCHER in the 7th - in line with the Network strategy that identified cogeneration of process heat and power as the main specific mission of HTR. HTR can indeed address the growing energy needs of industry presently fully relying on fossil fuel combustion with a CO2-lean generation technology, thanks to its high operating temperature and to its unique flexibility obtained from its large thermal inertia and its low power. Relying on the legacy of the former European leadership in HTR technology, this programme has addressed specific developments required for industrial process heat applications and for increasing HTR performances (higher temperatures and fuel burn-up). Decisive achievements have been obtained concerning fuel manufacturing and irradiation behaviour, key components and their materials, safety, computer code validation and specific HTR waste (fuel and graphite) management. Key experiments have been performed or are still ongoing: irradiation of graphite, fuel and vessel materials and the corresponding post-irradiation examinations, safety tests and isotopic analyses; thermal-hydraulic tests of an Intermediate Heat Exchanger mock-up in helium; air ingress experiments for a block type core, etc. Through Euratom participation in the Generation IV International Forum (GIF), these achievements contribute to international cooperation. HTR-TN strategy has been recently integrated by the 'Sustainable Nuclear Energy Technology Platform' (SNE-TP) as one of the 3 'pillars' of its global nuclear strategy. It is also in line with the orientations and the timing of the 'Strategic Energy Technology Plan (SET-Plan)' for the development of

  16. Development of Non-Metallic Fuel Elements for a High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    In connection with fuel element development work for the high-temperature gas-coolcd reactor of the Brown-Boveri/Krupp Reaktorbau G.m.b.H., two different fuel element concepts were considered and developed. In both cases the fuel element consists of a graphite ball of 6 cm in diam. which contains the fuel insert, a cylindrical pellet of about 20 mm in diam. and 16 mm in height. The two concepts differ in the type of the.fuel insert as well as in the preparation of the graphite ball. In the first concept the fuel insert consists of a mixture of UC2 and graphite which is prepared by blending U3O8 and graphite, pressing them into pellets and reacting the two components in a vacuum furnace at 1800oC. The atomic ratio of U : C is 1:45. Since this type of fuel pellet does not retain the fission products completely the surrounding graphite sphere had to be made impervious to fission products by impregnation in order to obtain a fission-product retaining element. Permeabilities of the order of 10-6cm2/s could be achieved. In the second concept the fuel insert consists of a solid solution of UC in ZrC and is coated with a layer of ZrC. The molar ratio of UC to ZrC is 1 : 20. The fuel pellet preparation was accomplished by the following procedure: UO2, ZrO2, and graphite were mixed and pressed into pellets. The pellets were reacted to the carbides. Ball milling of the carbides was followed by hot pressing at temperatures o f 2000oC. Densities of more than 95% of the theoretical density could be achieved. A full description of the preparation and of some physical properties of the fuel pellets is given in the paper. A sufficient fission gas retention behaviour of this type of fuel insert which allows it to be put into unimpregnated graphite balls is expected. Other advantages of this kind of fuel are discussed. (author)

  17. High temperature reactor: Driving force to convert CO2 to fuel - HTR2008-58132

    International Nuclear Information System (INIS)

    The rapidly increasing cost of petroleum products and uncertainty of long-term supply have prompted the U.S. military to aggressively pursue production of alternative fuels (synfuels) such as coal-to-liquids (CTL). U.S. Air Force is particularly active in this effort while the entire military is involved in simultaneously developing fuel specifications for alternative fuels that enable a single fuel for the entire battle space; all ground vehicles, aircraft and fuel cells. By limiting its focus on coal, tar sands and oil shale resources, the military risks violating federal law which requires the use of synfuels that have life cycle greenhouse gas emissions less than or equal to emissions from conventional petroleum fuels. A climate-friendly option would use a high temperature nuclear reactor to split water. The hydrogen (H2) would be used in the reverse water gas shift (RWGS) to react with carbon dioxide (CO2) to produce carbon monoxide (CO) and water. The oxygen (O2) would be fed into a supercritical (SC) coal furnace. The flue gas CO2 emissions would be stripped of impurities before reacting with H2 in a RWGS process. Resultant carbon monoxide (CO) is fed, with additional H2, (extra H2 needed to adjust the stoichiometry: 2 moles H2 to one mole CO) into a conventional Fischer-Tropsch synthesis (FTS) to produce a heavy wax which is cracked and isomerized and refined to Jet Propulsion 8 (JP-8) and Jet Propulsion 5 (JP-5) fuels. The entire process offers valuable carbon-offsets and multiple products that contribute to lower syn-fuel costs and to comply with the federal limitation imposed on syn-fuel purchases. While the entire process is not commercially available, component parts are being researched; their physical and chemical properties understood and some are state-of-the-art technologies. An international consortium should complete physical, chemical and economic flow sheets to determine the feasibility of this concept that, if pursued, has broad applications

  18. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  19. Use of a Nuclear High Temperature Gas Reactor in a Coal-To-Liquids Process

    International Nuclear Information System (INIS)

    AREVA's High Temperature Gas Reactor (HTGR) can potentially provide nuclear-generated, high-level heat to chemical process applications. The use of nuclear heat to help convert coal to liquid fuels is particularly attractive because of concerns about the future availability of petroleum for vehicle fuels. This report was commissioned to review the technical and economic aspects of how well this integration might actually work. The objective was to review coal liquefaction processes and propose one or more ways that nuclear process heat could be used to improve the overall process economics and performance. Shell's SCGP process was selected as the gasifier for the base case system. It operates in the range of 1250 to 1600 C to minimize the formation of tars, oil, and methane, while also maximizing the conversion of the coal's carbon to gas. Synthesis gas from this system is cooled, cleaned, reacted to produce the proper ratio of hydrogen to carbon monoxide and fed to a Fischer-Tropsch (FT) reaction and product upgrading system. The design coal-feed rate of 18,800 ton/day produces 26.000 barrels/day of FT products. Thermal energy at approximately 850 C from a HTGR does not directly integrate into this gasification process efficiently. However, it can be used to electrolyze water to make hydrogen and oxygen, both of which can be beneficially used in the gasification/FT process. These additions then allow carbon-containing streams of carbon dioxide and FT tail-gas to be recycled in the gasifier, greatly improving the overall carbon recovery and thereby producing more FT fuel for the same coal input. The final process configuration, scaled to make the same amount of product as the base case, requires only 5,800 ton/day of coal feed. Because it has a carbon utilization of 96.9%, the process produces almost no carbon dioxide byproduct Because the nuclear-assisted process requires six AREVA reactors to supply the heat, the capital cost is high. The conventional plant is

  20. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  1. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    Science.gov (United States)

    Zweibaum, Nicolas

    The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). (Abstract shortened by UMI.).

  2. Research and development on the relating technology of a multi-purpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In a multi-purpose high temperature gas-cooled reactor, a heat exchanger, fuel elements and heat-insulating material are in severe heat transfer condition due to the circulating helium gas of about 1,0000C and pressure of 40 kg/cm2. It is thus necessary to acquire ample experimental data on such components. Studies made so far on the fuel elements and graphite in Japan Atomic Energy Research Institute and First Atomic Power Industry Group are described: fuel and its production test, out-of-pile test, irradiation test and FP release; graphite and high temperature irradiation test and post-irradiation test, mechanical strength, high temperature corrosion and selection of graphite. (Mori, K.)

  3. Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2010-02-01

    This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

  4. Analysis of characteristics of different working fluids for gas turbine cycle with high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Gas turbine cycle with high temperature gas-cooled reactor is the main direction of nuclear energy generation, which is with the advantages in terms of the safety and economy. The thermal and physical properties of helium, nitrogen, carbon dioxide and the mixtures were compared and analyzed in this paper. Further more, the heat transfer coefficient, pressure loss and the stage number of turbo-machines have been also compared. Results indicate that taking the mixture of helium and carbon dioxide as the working fluid of gas turbine cycle with high temperature gas-cooled reactor can not only improve the heat transfer coefficient and decrease the stage number of turbo-machinery, but also can limit the pressure loss to a certain level. (authors)

  5. Gas-Liquid Mass Transfer in a Slurry Bubble Column Reactor under High Temperature andHigh Pressure

    Institute of Scientific and Technical Information of China (English)

    杨卫国; 王金福; 金涌

    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 kLa 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 kLa values for H2 and CO in liquid paraffln/solid particles slurry bubble column reactors.

  6. Program status of the high temperature reactor development in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The status of the HTR development program in the Federal Republic of Germany in 1984 is characterized by the beginning of a transition phase from a national program to a commercial program. In the last 20 years the HTR technology program was strongly, nearly completely supported by the Federal Government and the State Government of North-Rhine-Westfalia. Funding of the program up to now exceeded 5 billion DM. Within this framework it was possible to establish competent-reactor-system companies, to enable industries to supply HTR- specific components including fuel elements and nuclear graphites, to maintain the strong engagement of the national centre KFA Juelich in general R and D activities, to build and operate the AVR-plant for more than 16 years, to erect the demonstration plant THTR-300 now approaching completion and to build and operate many efficient test facilities. Thereby the HTR technology development achieved a stage of maturity which is not only considered to be most advanced, but is also ready now for commerical deployment. The assessment report which comprised both the fast breeder and the HTR development included all major impacts, such as history, status, prospects, benefits, industrial aspects and international developments of the technology. The program description is facilitated by distinguishing the five major program elements: AVR, THTR-300, THTR follow-up plant, nuclear process heat program, fuel cycle activities

  7. Modular high-temperature gas-cooled reactor short term thermal response to flow and reactivity transients

    International Nuclear Information System (INIS)

    The short-term thermal response of the modular high-temperature gas-cooled reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These include loss of forced circulation (LOFC) without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described

  8. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    International Nuclear Information System (INIS)

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments

  9. A Soft-Switching Inverter for High-Temperature Advanced Hybrid Electric Vehicle Traction Motor Drives

    Energy Technology Data Exchange (ETDEWEB)

    Lai, Jason [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Yu, Wensong [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Sun, Pengwei [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Leslie, Scott [Powerex, Inc., Harrison, OH (United States); Prusia, Duane [Powerex, Inc., Harrison, OH (United States); Arnet, Beat [Azure Dynamics, Oak Park, MI (United States); Smith, Chris [Azure Dynamics, Oak Park, MI (United States); Cogan, Art [Azure Dynamics, Oak Park, MI (United States)

    2012-03-31

    The state-of-the-art hybrid electric vehicles (HEVs) require the inverter cooling system to have a separate loop to avoid power semiconductor junction over temperatures because the engine coolant temperature of 105°C does not allow for much temperature rise in silicon devices. The proposed work is to develop an advanced soft-switching inverter that will eliminate the device switching loss and cut down the power loss so that the inverter can operate at high-temperature conditions while operating at high switching frequencies with small current ripple in low inductance based permanent magnet motors. The proposed tasks also include high-temperature packaging and thermal modeling and simulation to ensure the packaged module can operate at the desired temperature. The developed module will be integrated with the motor and vehicle controller for dynamometer and in-vehicle testing to prove its superiority. This report will describe the detailed technical design of the soft-switching inverters and their test results. The experiments were conducted both in module level for the module conduction and switching characteristics and in inverter level for its efficiency under inductive and dynamometer load conditions. The performance will be compared with the DOE original specification.

  10. Loss-of-water accident analysis of the pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The high pressure helium and water/steam are respectively used as the primary and secondary coolant for the pebble-bed modular high temperature gas-cooled reactor (HTGR). Loss-of-water accident is one of the typical design basis accident (DBA), which would be caused by malfunction or current failure of the feed water pump, as well as the false action of the feed water valve. During the loss-of-water accident, due to the loss of the secondary heat sink, the temperature and pressure of primary coolant will increase. Subsequently, the reactor scram will be triggered by the protective signal of the “high flow rate proportion of primary circuit and secondary circuit” or the “high core inlet helium temperature”. For this type of the accident, the earlier open of the safety valve of the primary circuit should be avoided by reactor design. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), with the coupled analysis code TINTE-BLAST, accidents with different slowdown rate of the feed water supply have been studied. The important parameters, including the reactor power, fuel element temperature, inlet/outlet helium temperature of the core, and especially the primary pressure, are analyzed. The consequences with first scram signal succeeding or failing are compared. The results can prove that, according to the current design of the protection system, this kind of accident can be detected in time. The scram signal will trigger the reactor shut down quickly, without causing the earlier open of the safety valve. After the reactor is successfully shut down, due to the inherent safety feature of the HTGR, the temperature and the pressure in the primary circuit will increase very slowly. The temperature of the fuel element, as well as that of the components, will not exceed the design limitations. (author)

  11. Production of advanced materials by methods of self-propagating high-temperature synthesis

    CERN Document Server

    Tavadze, Giorgi F

    2013-01-01

    This translation from the original Russian book outlines the production of a variety of materials by methods of self-propagating high-temperature synthesis (SHS). The types of materials discussed include: hard, refractory, corrosion and wear-resistant materials, as well as other advanced and speciality materials. The authors address the issue of optimal parameters for SHS reactions occurring during processes involving a preliminary metallothermic reduction stage, and they calculate this using thermodynamic approaches. In order to confirm the effectiveness of this approach, the authors describe experiments focussing on the synthesis of elemental crysalline boron, boron carbides and nitrides. Other parts of this brief include theoretical and experimental results on single-stage production of hard alloys on the basis of titanium and zirconium borides, as well as macrokinetics of degassing and compaciton of SHS-products.This brief is suitable for academics, as well as those working in industrial manufacturing com...

  12. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

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

    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; Strydom, Gerhard [Idaho National Laboratory

    2016-03-01

    • Provide an initial summary description of the design and its main attributes o Summarize the main Test Reactor attributes: reactor type, power, coolant, irradiation conditions (fast and thermal flux levels, number of test loops, positions and volumes), costs (project, operational), schedule and availability factor. o Identify secondary missions and power conversion options, if applicable. o Include statements on the envisioned attractiveness of the reactor type in relation to anticipated domestic and global irradiation services needs, citing past and current trends in reactor development and deployment. o Include statements on Test Reactor scalability (e.g. trade-off between size, power/flux levels and costs), prototypical conditions, overall technology maturity of the specific design and the general technology type. The intention is that this summary must be readable as a stand-alone section.

  14. Current hurdles to the success of high-temperature membrane reactors

    NARCIS (Netherlands)

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

    1994-01-01

    High-temperature catalytic processes performed using inorganic membranes have been in recent years a fast growing area of research, which seems to have not yet reached its peak. Chemical engineers, catalysts and materials scientists have addressed this topic from different viewpoints in a common eff

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

    NARCIS (Netherlands)

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

    2011-01-01

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

  16. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    Science.gov (United States)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  17. US/FRG joint report on the pebble bed high temperature reactor resource conservation potential and associated fuel cycle costs

    International Nuclear Information System (INIS)

    Independent analyses at ORNL and KFA have led to the general conclusion that the flexibility in design and operation of a high-temperature gas-cooled pebble-bed reactor (PBR) can result in favorable ore utilization and fuel costs in comparison with other reactor types, in particular, with light-water reactors (LWRs). Fuel reprocessign and recycle show considerable promise for reducing ore consumption, and even the PBR throwaway cycle is competitive with fuel recycle in an LWR. The best performance results from the use of highly enriched fuel. Proliferation-resistant measures can be taken using medium-enriched fuel at a modest ore penalty, while use of low-enriched fuel would incur further ore penalty. Breeding is possible but net generation of fuel at a significant rate would be expensive, becoming more feasible as ore costs increase substantially. The 233U inventory for a breeder could be produced by prebreeders using 235U fuel

  18. Preliminary studies of coolant by-pass flows in a prismatic very high temperature reactor using computational fluid dynamics

    International Nuclear Information System (INIS)

    Three dimensional computational fluid dynamic (CFD) calculations for a 1/12 sector of a prismatic block through the core of a prismatic very high temperature gas-cooled reactor (VHTR) were conducted to investigate the influence of gap geometry on flow and temperature distributions in the reactor core using commercial CFD code FLUENT. Parametric calculations changing the gap width in a whole core length model of fuel and reflector columns were performed. The simulations show the effects of core by-pass flows in the heated core region by comparing results for several gap widths including zero gap width. The calculation results underline the importance of considering inter-column gap width for the evaluation of maximum fuel temperatures and temperature gradients in fuel blocks. It is shown that temperatures of core outlet flow from gaps and channels are strongly affected by the gap width of by-pass flow in the reactor core. (author)

  19. Preliminary studies of coolant by-pass flows in a prismatic very high temperature reactor using computational fluid dynamics

    International Nuclear Information System (INIS)

    Three dimensional computational fluid dynamic (CFD) calculations of a typical prismatic very high temperature gas-cooled reactor (VHTR) were conducted to investigate the influence of gap geometry on flow and temperature distributions in the reactor core using commercial CFD code FLUENT. Parametric calculations changing the gap width in a whole core length model of fuel and reflector columns were performed. The simulations show the effects of core by-pass flows in the heated core region by comparing results for several gap widths including zero gap width. The calculation results underline the importance of considering inter-column gap width for the evaluation of maximum fuel temperatures and temperature gradients in fuel blocks. In addition, it is shown that temperatures of core outlet flow from gaps and channels are strongly affected by the gap width of by-pass flow in the reactor core.

  20. Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230, and Model Alloys

    International Nuclear Information System (INIS)

    The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHXs) of (very) high temperature reactors ((V)-HTRs). The behavior under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra- and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. The alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow (Rouillard, F., 2007, 'Mecanismes de formation et de destruction de la couche d'oxyde sur un alliage chrominoformeur en milieu HTR, Ph.D. thesis, Ecole des Mines de Saint-Etienne, France). To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T approximate to 1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617, and model alloys 1178, 1181, and 1201. This coupling makes it possible for the thermodynamic equilibrium to be obtained between the vapor phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared with that of the pure substance (Cr) at the same temperature

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

    International Nuclear Information System (INIS)

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

  2. Application of TL dosemeters for dose distribution measurements at high temperatures in nuclear reactors.

    Science.gov (United States)

    Osvay, M; Deme, S

    2006-01-01

    Al2O3:Mg,Y ceramic thermoluminescence dosemeters were developed at the Institute of Isotopes for high dose applications at room temperatures. The glow curve of Al2O3:Mg,Y exhibits two peaks--one at 250 degrees C (I) and another peak at approximately 400 degrees C (II). In order to extend the application of these dosemeters to high temperatures, the effect of irradiation temperature was investigated using temperature controlled heating system during high dose irradiation at various temperatures (20-100 degrees C). The new calibration and measuring method has been successfully applied for dose mapping within the hermetic zone of the Paks Nuclear Power Plant even at high temperature parts of blocks.

  3. Mechanical problems in high-temperature gas-cooled reactor structures

    International Nuclear Information System (INIS)

    Graphite is adopted as the major structural material in HTGR so the study of mechanical behaviour of graphite under high temperature, high radiation and oxidization becomes a main point. Physical and mechanical properties of graphite, fatigue behaviour of graphite and stress classification and failure criteria under the surroundings in question are discussed. The paper also gives an introduction to creeping and cracking problems of concrete in the stress analysis of HTGR pressure vessels (PCPV)

  4. Investigation of chemical characteristics of primary helium gas coolant of HTTR (high temperature engineering test reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hamamoto, Shimpei, E-mail: hamamoto.shimpei@jaea.go.jp [HTTR Operation Section, Department of HTTR, Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan); Shimazaki, Yosuke [HTTR Reactor Engineering Section, Department of HTTR, Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan); Furusawa, Takayuki; Nemoto, Takahiro; Inoi, Hiroyuki [HTTR Operation Section, Department of HTTR, Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan); Takada, Shoji, E-mail: takada.shoji@jaea.go.jp [HTTR Reactor Engineering Section, Department of HTTR, Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan)

    2014-05-01

    The technical basis of helium gas purification control for HTGRs was established by verifying the design of the Primary Helium Purification System (PHPS) of the HTTR by showing that the measured concentrations of impurities of the primary helium coolant were restricted below the criteria of control to protect the graphite oxidation, and that the carburization atmosphere was maintained to keep intact of metallic high temperature components, in the 30-day continuous operation and the 50-day long-term high temperature operation. The analytical model, which was newly established by improving the conventional method that predicted the impurity concentrations conservatively higher than the measured values, predicted the composition of the impurities such as H{sub 2}, CO, H{sub 2}O and CO{sub 2}, which is determined by the temperature dependency of release of impurities during the rated power operation adequately. In contrast, it was revealed that the measured concentration of H{sub 2}O remarkably decreased while the concentration of CO increased in the primary helium coolant in the long-term high temperature operation.

  5. Investigation of chemical characteristics of primary helium gas coolant of HTTR (high temperature engineering test reactor)

    International Nuclear Information System (INIS)

    The technical basis of helium gas purification control for HTGRs was established by verifying the design of the Primary Helium Purification System (PHPS) of the HTTR by showing that the measured concentrations of impurities of the primary helium coolant were restricted below the criteria of control to protect the graphite oxidation, and that the carburization atmosphere was maintained to keep intact of metallic high temperature components, in the 30-day continuous operation and the 50-day long-term high temperature operation. The analytical model, which was newly established by improving the conventional method that predicted the impurity concentrations conservatively higher than the measured values, predicted the composition of the impurities such as H2, CO, H2O and CO2, which is determined by the temperature dependency of release of impurities during the rated power operation adequately. In contrast, it was revealed that the measured concentration of H2O remarkably decreased while the concentration of CO increased in the primary helium coolant in the long-term high temperature operation

  6. Long-term high temperature fatigue properties of new structural materials for nuclear reactors

    International Nuclear Information System (INIS)

    This study aims to evaluate fatigue strength properties at long-term high temperature of 316FR steels developed for structural materials applied to high temperature components in the 21st Century such as Demonstration FBR, and to obtain design indicators on high temperature strength properties arranged on chemical composition and grain size on a base of 316FR stainless steels. As results obtained by the study, it could be found that as a result of systematic examinations on temperatures and strain rate dependence on symmetric triangular strain waveform of 316FR steels between 500 to 600degC, temperature of fatigue life, and strain rate dependence can be integrally evaluated by a parameter analysis method developed by authors and life prediction at ultra low strain rate of less than 10-6/s and its experimental verification could be carried out. And, it was also found that as a result of evaluation of creep fatigue life and creep rupture properties on materials prepared by controlling main elements such as carbon, nitrogen, manganese, chromium, and performing grain refining through thermomechanical treatment under aiming to further upgrade creep fatigue life properties of 316FR steels, it was clarified to enable to carry out about three times of life extension by combined effects of low carbon (0.002 wt-%) and grain refining (7 in grain number). (G.K.)

  7. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  8. Modeling the high-temperature gas-cooled reactor process heat plant: a nuclear to chemical conversion process

    International Nuclear Information System (INIS)

    The high-temperature heat available from the High-Temperature Gas-Cooled Reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design

  9. Fission product transport in the high temperature gas-cooled reactor: Theory, program development and verification by recalculation of experiments

    International Nuclear Information System (INIS)

    The high temperature gascooled reactor (HTGR) reaches a special standard in safety because of its high temperature resistent fuel element. After all the possibility of fission product releases can not be excluded without further investigations for HTGRs. The mechanisms of fission product releases, which occur in case of such hypothetical events, are the subject of this work. The main focus of the investigation is how the fission products, which have been released, are re-adsorpted and prevented through this mechanism from being released in the environment. A strong effect of re-adsorption is expected, because experiments have shown that graphite, which is 100% of the core material, has an excellent capability to hold back fission products. With the program tools developed to calculate the fission product transport mechanisms, the corresponding experiments are recalculated and also fission product release calculations are carried out. (orig./HP)

  10. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems.

    Science.gov (United States)

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes. PMID:26520986

  11. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  12. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

    Science.gov (United States)

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  13. ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

    International Nuclear Information System (INIS)

    ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

  14. Dragon project reference design assessment study for a 528 MW (E) thorium cycle high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The report presents an assessment of the feasibility, safety and cost of a large nuclear power station employing a high temperature gas-cooled reactor. A thermal output 1250 MW was chosen for the study, resulting in a net electrical output of 528.34 MW from a single reactor station, or 1056.7 MW from a twin reactor station. A reference design has been developed and is described. The reactor uses a U-235/Th-232/U-233 fuel cycle, on a feed and breed basis. It is believed that such a reactor could be built at an early date, requiring only a relatively modest development programme. Building costs are estimated to be Pound46.66/kW for a single unit station and Pound42.6/kW for a twin station, with power generation costs of 1.67p/kWh and 1.50p/kWh respectively. Optimisation studies have not been carried out and it should be possible to improve on the costs. The design has been made as flexible as possible to allow units of smaller or larger outputs to be designed with a minimum of change. (U.K.)

  15. High temperature gas-cooled pebble bed reactor steady state thermal-hydraulics analyses based on CFD method

    International Nuclear Information System (INIS)

    Background: Based on general purpose CFD code Fluent, the PBMR-400 full load nominal condition thermal-hydraulics performance was studied by applying local thermal non-equilibrium porous media model. Purpose: In thermal hydraulics study of the gas cooled pebble bed reactor, the core of the reactor can be treated as macroscopic porous media with strong inner heat source, and the original Fluent code can not handle it properly. Methods: By introducing a UDS in the calculation domain of the reactor core and subjoining a new resistance term, we develop a non-equilibrium porous media model which can give an accurate description of the core of the pebble bed. The mesh of CFD code is finer than that of the traditional pebble bed reactor thermal hydraulics analysis code such as THERMIX and TINTE, thus more information about coolant velocity fields, temperature field and solid phase temperature field can be acquired. Results: The nominal condition calculation results of the CFD code are compared to those of the well-established thermal-hydraulic code THERMIX and TINTE, and show a good consistency. Conclusion: The extended local thermal non-equilibrium model can be used to analyse thermal-hydraulics of high temperature pebble bed type reactor. (authors)

  16. Design criteria, production and total integrity assessment of fuels of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report describes the design criteria, production and total integrity of the HTTR fuels for the safety design of the reactor. The fuels were designed so that they should not lose their integrity even though taking account of various kinds of possible deteriorations during reactor service. Sufficiently low values of initial (as-produced) fuel failure fractions have been achieved, and experience of fuel production is enough for full core loading. Results of the present assessment have shown that total integrity of the fuels will be maintained successfully in terms of coating failure of the fuel particles, thermal and mechanical performance of the fuel compacts, graphite sleeves and fuel assemblies. (author)

  17. Thermodynamic analysis of performance of steam methane reforming hydrogen production system connected with high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Thermodynamic analysis of performance of steam methane reforming hydrogen production system connected with High-Temperature Gas-Cooled Reactor is presented, which provides a framework for further detailed research. Complete reaction model and equilibrium reaction model were developed. System efficiency and hydrogen output variation related to process parameters were researched. Limit value of performance index and optimum process parameter were determined. The comparison of equilibrium reaction model prediction to experimental data shows that the equilibrium reaction model is appropriate for preliminary analysis for the system. (authors)

  18. Numerical analysis of steam reformer of steam methane reforming hydrogen production system connected with high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    In order to quantitatively analyze the performance of the helium-heated reformer used in steam methane reforming hydrogen production system connected with high temperature gas cooled reactor, a dynamic model has been set up based on one-dimension quasi-homogeneous phase model. And a computer program is development. Model verification is performed under steady state using test results of Japan Atomic Energy Institute. The steady state calculation results fit well with the experiment results. Reaction velocity is not the main factor influencing the performance. Reformer tube with finned central tube improves the performance remarkably comparing with smooth central tube. (authors)

  19. Initial study on steam reformer of high-temperature gas-cooled reactor powered steam methane reforming hydrogen production system

    International Nuclear Information System (INIS)

    Based on one-dimension quasi-homogeneous phase model, a dynamic model for single-tube steam reformer of high-temperature gas-cooled reactor was presented, and computer program was developed. Steady state calculation and analysis were performed for the steam reformer design by Japan Atomic Energy Research Institute. The results show that heat loss at the entrance of helium influences the steam reformer performance remarkably, and reaction velocity is not main factor influencing the performance. The steady state calculation results fit well with experiment results. (authors)

  20. Suspension of a steam raising unit in a reinforced concrete pressure vessel of a high temperature reactor

    International Nuclear Information System (INIS)

    The invention concerns a supporting device in a steam raising unit, which is situated in the reinforced concrete pressure vessel of a high temperature reactor. The supporting device is situated in the space between the steam raising unit skirt and the steam raising unit jacket. If the suspension of the steam raising unit skirt fails, the steam raising unit skirt is well fixed by the supporting device in the steam raising unit jacket, without the feed water and steam pipes in the upper part of the steam raising unit being damaged. (orig.)

  1. Preliminary study on rotor dynamics of magnetic bearing for 10MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The finite element method and MSC. Marc software are applied to analyze the rotor modal of magnetic bearings for power conversion unit (PCU) of 10 MW high temperature gas-cooled reactor (HTR-10). The effects of the magnetic bearings sustaining stiffness on the rotor natural frequencies were studied. Results show that the natural frequencies may be adjusted by changing the sustaining stiffness and rotor material. It is very important for the magnetic bearing to pass two order bending natural frequencies and design control system

  2. High and very high temperature reactor research for multipurpose energy applications (RAPHAEL)

    International Nuclear Information System (INIS)

    The sections of the contribution are as follows: Achievements and future R and D needs in baseline technologies (Fuel and fuel cycle; Materials and components - reactor vessel, intermediate heat exchanger, graphite internals; Safety; Computer code qualification; Waste management; Education and training; International dimension of the European R and D); Towards application; and The path forward (P.A.)

  3. Conceptual design of a passive, inherently safe emergency shutdown rod for high-temperature reactor applications

    International Nuclear Information System (INIS)

    The concept of a passive, inherently safe, and fail-safe design for an emergency control rod is presented. The functioning of the rod is based solely on inexorable physical laws. The operation of the rod in its emergency function does not require the intervention of a human operator, nor does it rely on any signal from a monitoring or safety system. Although the concept could be applicable to a variety of reactors (provided a normal temperature range is specified), in this paper, the concept is applied to the emergency shutdown of a pebble-bed reactor. The preliminary study presented here demonstrates that the proposed Electro-Magnetic Optimally Scramming Control Rod (EM-OSCR) naturally operates when needed. The rod is held out of the core region by the force of an electromagnet. The force is generated by a current carried by a conductor, a portion of which passes near or through the reactor core region. When the temperature in the conductor increases because of an increase in temperature in the reactor, the conductor resistivity increases. This, in turn, leads to a current decrease. When the current decreases below the level necessary to hold the rod up, the rod is released and it falls into the core under the effect of gravity. (author)

  4. The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 4

    International Nuclear Information System (INIS)

    In this volume the design conditions for a helium-heated steam reformer in the primary circuit of a high-temperature reactor are explained as far as today's knowledge allows. For the realization of helium-heated steam reformers, some fundamental questions at first occur regarding the heating temperature, heat fluxes, suitable materials and design solutions for steam reformers. It is shown that following the development program carried out until now, solutions to these questions can be seen. Moreover, details are given about the heat transfer, the mechanical design and the behaviour of reformer materials in helium with regard to H2- and T-permeation as well as corrosion. Furthermore, questions about the choice of the lay-out data, the design form, the arrangement in the helium circuits of the nuclear reactor and the necessary development steps are handled. Some design examples of heat exchangers for a 3,000 MW(th)-plant are given, too. (orig.)

  5. Computation of fission product distribution in core and primary circuit of a high temperature reactor during normal operation

    International Nuclear Information System (INIS)

    The fission product release during normal operation from the core of a high temperature reactor is well known to be very low. A HTR-Modul-reactor with a reduced power of 170 MWth is examined under the aspect whether the contamination with Cs-137 as most important nuclide will be so low that a helium turbine in the primary circuit is possible. The program SPTRAN is the tool for the computations and siumlations of fission product transport in HTRs. The program initially developed for computations of accident events has been enlarged for computing the fission product transport under the conditions of normal operation. The theoretical basis, the used programs and data basis are presented followed by the results of the computations. These results are explained and discussed; moreover the consequences and future possibilities of development are shown. (orig./HP)

  6. A New Class of Functionally Graded Cearamic-Metal Composites for Next Generation Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 C. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with various microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.

  7. "A New Class od Functionally Graded Cearamic-Metal Composites for Next Generation Very High Temperature Reactors"

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose; Mrs. Judith Maro, Nuclear Reactor Laboratory, MIT

    2008-05-01

    Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with various microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.

  8. Chemical compatibility of SiC composite structures with fusion reactor helium coolant at high temperatures

    International Nuclear Information System (INIS)

    The thermodynamic stability of SiC/SiC composite structures proposed for fusion applications is presented in this paper. Minimization of the free energy for reacting species in the temperature range 773-1273 K is achieved by utilizing the NASA-Lewis Chemical Equilibrium Thermodynamics Code (CET). The chemical stability of the matrix (SiC), as well as several fiber coatings (BN and graphite) are studied. Helium coolant is assumed to contain O2 and water moisture impurities in the range 100-1000 ppm. The work is applied to recent Magnetic and Inertial Confinement Conceptual designs. The present study indicates that the upper useful temperature limit for SiC/SiC composites, from the standpoint of high-temperature corrosion, will be in the neighborhood of 1273 K. Up to this temperature, corrosion of SiC is shown to be negligible. The main mechanism of weight loss will be by evaporation to the plasma side. The presence of a protective SiO2 condensed phase is discussed, and is shown to result in further reduction of high-temperature corrosion. The thermodynamic stability of C and BN is shown to be very poor under typical fusion reaction conditions. Further development of chemically stable interface materials is required. (orig.)

  9. Review of activities of Research Association of High Temperature Gas Cooled Reactor Plant (RAHP)

    International Nuclear Information System (INIS)

    The Research Association of the HTGR Plant (RAHP) is the sole research association in the private or industrial sector of Japan with respect to HTGR Plants. It was established in 1985, composing of professors, representatives of electric power companies, and fabricators of nuclear plant and fuels Activities in these years were to analyze world trends of R and D, to identify techno-economical issues to be cleared, to set-up fundamental development strategies, and to put the results of the studies into actions towards commercialization of the HTGR. Conclusions obtained through the activities so far are: (1) From the view point of effective use of energy and reduction of environmental impacts on a global scale, development of nuclear power is essential, in particular of the HTGR, because of its very highly inherent safety and feasibility of high temperature heat uses. The role of the HTGR is inter-complementary with those of LWR and FBR; (2) Future subjects on the HTGR are technical demonstration of its unique characteristics, economic prospects, public acceptance (PA) and industrial acceptance, R and D through international cooperation and share in role, and successful realization of demonstration plant(s). RAHP is to start a survey on HTGR from nuclear fuel cycle point of view to have a better outlook on future needs of high temperature heat uses. 6 refs

  10. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  11. Development, Implementation and Application of Micromechanical Analysis Tools for Advanced High Temperature Composites

    Science.gov (United States)

    2005-01-01

    This document contains the final report to the NASA Glenn Research Center (GRC) for the research project entitled Development, Implementation, and Application of Micromechanical Analysis Tools for Advanced High-Temperature Composites. The research supporting this initiative has been conducted by Dr. Brett A. Bednarcyk, a Senior Scientist at OM in Brookpark, Ohio from the period of August 1998 to March 2005. Most of the work summarized herein involved development, implementation, and application of enhancements and new capabilities for NASA GRC's Micromechanics Analysis Code with Generalized Method of Cells (MAC/GMC) software package. When the project began, this software was at a low TRL (3-4) and at release version 2.0. Due to this project, the TRL of MAC/GMC has been raised to 7 and two new versions (3.0 and 4.0) have been released. The most important accomplishments with respect to MAC/GMC are: (1) A multi-scale framework has been built around the software, enabling coupled design and analysis from the global structure scale down to the micro fiber-matrix scale; (2) The software has been expanded to analyze smart materials; (3) State-of-the-art micromechanics theories have been implemented and validated within the code; (4) The damage, failure, and lifing capabilities of the code have been expanded from a very limited state to a vast degree of functionality and utility; and (5) The user flexibility of the code has been significantly enhanced. MAC/GMC is now the premier code for design and analysis of advanced composite and smart materials. It is a candidate for the 2005 NASA Software of the Year Award. The work completed over the course of the project is summarized below on a year by year basis. All publications resulting from the project are listed at the end of this report.

  12. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

  13. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hawari, Ayman [North Carolina State Univ., Raleigh, NC (United States). Dept. of Nuclear Engineering; Ougouag, Abderrafi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-07-08

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  14. Joining of Oxide Dispersion Strengthened Steels for Advanced Reactors

    Science.gov (United States)

    Baker, B. W.; Brewer, L. N.

    2014-12-01

    The design, manufacture, and experimental analysis of structural materials capable of operation in the high temperatures, corrosive environments, and radiation damage spectra of future reactor designs remain one of the key pacing items for advanced reactor designs. The most promising candidate structural materials are vanadium-based refractory alloys, silicon carbide composites and oxide dispersion strengthened steels. Of these, oxide dispersion strengthened steels are a likely near-term candidate to meet required demands. This paper reviews different variants of oxide dispersion strengthened steels and discusses their capability with regard to high-temperature strength, corrosion resistance, and radiation damage resistance. Additionally, joining of oxide dispersion strengthened steels, which has been cited as a limiting factor preventing their use, is addressed and reviewed. Specifically, friction stir welding of these steels is reviewed as a promising joining method for oxide dispersion strengthened steels.

  15. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  16. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials

    International Nuclear Information System (INIS)

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  17. A simplified Probabilistic Safety Assesment of a Steam-Methane Reforming Hydrogen Production Plant coupled to a High-Temperature Gas Cooled Nuclear Reactor

    OpenAIRE

    Nelson Edelstein, Pamela; Flores Flores, Alain; Francois Lacouture, Juan Luis

    2005-01-01

    A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute’s (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP...

  18. The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

    2014-08-15

    To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

  19. Effect of nitrogen on high temperature mechanical properties of type 316LN SS for fast reactor applications

    International Nuclear Information System (INIS)

    Long term creep, low cycle fatigue and creep-fatigue interaction properties as well as compatibility with liquid sodium coolant, govern the choice of materials for out-of-core structural components of sodium cooled fast reactors (SFRs). 316 L(N) SS containing 0.07 Wt.% nitrogen is the current preferred material for all the high temperature structural components of SFRs. For the prototype Fast Breeder Reactor also, which is nearing commissioning at Kalpakkam, 316 LN (SS) has been used for all the major high temperature structural components. It is proposed to design future SFRs for a life of at least 60 years. Therefore materials with higher creep and low cycle fatigue strength are required. Increasing the nitrogen content in 316L(N) SS is one of the approaches being pursued at Indira Gandhi Centre for Atomic Research (IGCAR) to develop a SFR structural material suitable for 60 years design life. Towards this, a strong R and D programme is currently underway at IGCAR. The effect of nitrogen in the range of 0.07 to 0.22 Wt.% on the tensile creep and low cycle fatigue behavior of 316 LN SS has been extensively investigated. Both tensile and creep strength were found to increase with increase in nitrogen content. High temperature low cycle fatigue life was found to peak in 316LN SS containing 0.14 Wt.%. The paper discusses the monotonic and cyclic deformation and fracture mechanisms by which nitrogen improves the creep and low cycle fatigue properties of 316 LN SS. (author)

  20. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  1. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  2. An assessment of the modular HTGR [High Temperature Gas-Cooled Reactor] containment system

    International Nuclear Information System (INIS)

    Utility/user requirements for the MHTGR provide extremely stringent criteria for radionuclide release to the environment. In response to these top-level requirements, the MHTGR has evolved as a design emphasizing passive safety features. A crucial element of the MHTGR safety concept is the containment system which places primary emphasis on retention of fission products in the fuel. In addition to the particle-coating containment, the primary system pressure boundary and reactor building designs contribute to radionuclide retention. The philosophy of the MHTGR containment system is discussed in this paper and an assessment of its effectiveness is provided

  3. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  4. The challenge of introducing high-temperature reactor plants onto the international power plant market

    International Nuclear Information System (INIS)

    Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

  5. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-11-01

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  6. The pebble bed high temperature reactor as a source of nuclear process heat. Vol. 2

    International Nuclear Information System (INIS)

    A theoretical analysis is given for a series of 8 different variants of the pebble-bed reactor in the 'once through' fuel management scheme. The comparison gives some insight into the parametric sensitivities and into the development potential of this type. The thorium/U-233 recycling fuel cycle allows to increase the conversion ratio up to the range between 0.90 and 0.95. The feasibility for a changeover between different fuel cycles under full power operation. - The study is complemented by a review of the relevant previous investigations. (orig.)

  7. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  8. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  9. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  10. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  11. High temperature electronics

    Science.gov (United States)

    Seng, Gary T.

    1991-03-01

    In recent years, the aerospace propulsion and space power communities have acknowledged a growing need for electronic devices that are capable of sustained high-temperature operation. Aeropropulsion applications for high-temperature electronic devices include engine ground test instrumentation such as multiplexers, analog-to-digital converters, and telemetry systems capable of withstanding hot section engine temperatures in excess of 600 C. Uncooled operation of control and condition monitoring systems in advanced supersonic aircraft would subject the electronics to temperatures in excess of 300 C. Similarly, engine-mounted integrated electronic sensors could reach temperatures which exceed 500 C. In addition to aeronautics, there are many other areas that could benefit from the existence of high-temperature electronic devices. Space applications include power electronic devices for space platforms and satellites. Since power electronics require radiators to shed waste heat, electronic devices that operate at higher temperatures would allow a reduction in radiator size. Terrestrial applications include deep-well drilling instrumentation, high power electronics, and nuclear reactor instrumentation and control. To meet the needs of the applications mentioned previously, the high-temperature electronics (HTE) program at the Lewis Research Center is developing silicon carbide (SiC) as a high-temperature semiconductor material. Research is focused on developing the crystal growth, growth modeling, characterization, and device fabrication technologies necessary to produce a family of SiC devices. Interest in SiC has grown dramatically in recent years due to solid advances in the technology. Much research remains to be performed, but SiC appears ready to emerge as a useful semiconductor material.

  12. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    International Nuclear Information System (INIS)

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes

  13. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  14. Developments in advanced high temperature disc and blade materials for aero-engine gas turbine applications

    OpenAIRE

    Everitt, S

    2012-01-01

    The research carried out as part of this EngD is aimed at understanding the high temperature materials used in modern gas turbine applications and providing QinetiQ with the information required to assess component performance in new propulsion systems. Performance gains are achieved through increased turbine gas temperatures which lead to hotter turbine disc rims and blades. The work has focussed on two key areas: (1) Disc Alloy Assessment of High Temperature Properties; and (2) Thermal Barr...

  15. GT-MHR international project of high-temperature helium cooled reactor with direct gas-turbine power conversion cycle

    International Nuclear Information System (INIS)

    Full text: The international project of gas-turbine modular helium-cooled reactor (GT-MHR) presented in the report is a realization of high-temperature technology and is based on the experience with helium-cooled reactors with ceramic fuel particles and on innovative solutions concerning power conversion system with closed gas-turbine cycle and turbomachine with electromagnetic bearings. The international GT-MHR Project is currently being jointly developed by USA and Russia for disposition of excessive weapon-grade plutonium. For commercial electricity generation, the reactor will use uranium fuel. The GT-MHR combines a gas-cooled modular helium reactor (MHR) and a highly efficient integrated gas-turbine power conversion system (Brayton cycle) with expected cycle efficiency up to 48%. The reactor and power conversion unit are located in an underground concrete silo. The GT-MHR technical characteristics and design features assure: high level of passive safety that completely prevents core melting in accidents with any scenario, including full loss of inert coolant; low level of thermal and radiation impact to the environment; capability to use various fuels in the core (e.g. low-enriched uranium, mixed uranium-thorium and uranium-plutonium fuel, or plutonium fuel) without modifying the core design; meeting the non-proliferation requirements through technology and properties of ceramic fuel particles; capability to achieve coolant temperatures of up to 1000 deg. C, which is needed for various industrial processes; high efficiency of electricity generation. A whole complex of research and development activities is being carried out by RRC KI, OKBM, VNIINM, 'Lutch', and other Russian organizations in order to support key design solutions, primarily on fuel, turbomachine with electromagnetic bearings, structural materials, vessels and computer codes. At present, the GT-MHR capability to generate high-grade heat at temperatures up to 1000 deg. C makes it the only

  16. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  17. Design and energy analysis of a electrolytic hydrogen production process by means of a high temperature nuclear reactor; Diseno y analisis energetico de un proceso de produccion de hidrogeno electrolitico por medio de un reactor nuclear de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2008-07-01

    In this work an energy analysis to a process of production of hydrogen by means of electrolysis of high temperature is realized. This electrolysis type, unlike conventional electrolysis allows us to reach efficiencies of up to 60% because when increasing the temperature of the water, providing to its thermal energy, diminishes the demand of electrical energy required to separate the molecule of the water. Nevertheless, to obtain these efficiencies it is needed to have superheated aqueous vapor to but of 850 centigrade degrees, temperatures that can be reached about high temperature reactor; HTGR. In the present work it is mentioned to introduction way the importance of the hydrogen like energy vector and the advantages of obtaining it by means of nuclear energy. The electrolysis process of high temperature is described and a design is realized of this from its coupling to a nuclear power plant PBMR. The technological advances on which it counts the PBMR; efficiencies of 48% for optimized plants, their modular design and the thermodynamic cycle recuperative Brayton where upon operate; make the short term ideal candidate for the production of hydrogen. The thermodynamic analysis of optimized plant PBMR appears in another work, here the results of the balance of mass and energy involved in the process appear of hydrogen generation and the complete analysis of this. The result is a complete model of generation of hydrogen by electrolysis of high temperature coupled to an optimized plant PBMR that will be implemented for its dynamic simulation later. (Author)

  18. Application of high-temperature reactors in iron and steel industry with consideration of influences on location and environment

    International Nuclear Information System (INIS)

    High temperature reactors (HTR) are able to offer the various sorts of energy necessary for sponge iron technology (in a combined system of iron ore reduction and electric arc furnace) such as: 1) process heat for generating reduction gas from fossil raw materials or nuclear water splitting and 2) steam for electricity generating purposes. For the present the most promising model of combining HTR and iron and steel works seems to be the non integrated application. The locations of the reactors depend on the optimal location for reduction gas generation. The gas will be delivered to the metallurgical plants through a piping system. The electricity produced by the HTR could be fed into the local electrical grid. This article assumed an existing hydrogen distribution system and the possibility of direct reduction with hydrogen. For the direct reduction process a shaft furnace is chosen. The economy of reactor application is investigated and future potential use till the year 2,000 is estimated. Besides that the effects of HTR application on location and environment of iron and steel industry are described. (orig.)

  19. Basic studies on high-temperature engineering

    International Nuclear Information System (INIS)

    In response to increasing interest in high-temperature, gas-cooled reactors (HTGRs) and the need for improved knowledge of materials for nuclear applications that resist high temperatures, the NEA organised a first information exchange meeting on basic studies in the field of high-temperature engineering.The meeting was held in Paris on 27-29 september 1999 with 50 participants from 12 countries and four international organisations. Thirty two papers were submitted. The proceedings of the meeting cover studies on irradiation effects on advanced materials, safety-related behaviour of HTGRs and in-pile reactor instrumentation development. They also include recommendations for further promotion of international collaboration

  20. Steam generator materials performance in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 7600C and produce superheated steam at 5380C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 106 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

  1. Gas Turbine High Temperature Gas (Helium) Reactor Using Pebble Bed Fuel Derived from Spent Fuel

    International Nuclear Information System (INIS)

    Project goals: Build on the $1B investment spent during the NGNP Project for the only true Inherently Safe Small Modular Reactor Design – the only SMR design that can make this claim due to negative temperature coefficient of reactivity - no containment required – less construction cost. NPMC in Partnership with Pebble Bed Modular Group, a fully owned subsidiary of Eskom, RSA to Factory Build Complete Plant in Modular Sections at Factory Site in Oswego, NY for transport to site by rail or shipping for world wide export. NPMC will provide Project and Construction Management of all new builds from plant sites through construction, commissioning and startup using local labor. License and Construct ion of spent fuel processing facility in both NY and South Africa using Proven Technology. Ultimate goals of project: 1. Award of the 2013 US DOE Innovative SMR $452M cost share grant for US NRC License Certification 2.Build Full Scale Demonstration Plant at Koeburg, RSA with World Bank Funding managed by NPMC in collaboration with our legal firm, Haynes and Boone LLP 3. Take Plant Orders Immediately (10% Down Payment) 4. Form Strategic Alliance with Domestic and/or International Utility

  2. 10MW高温气冷实验堆的堆体结构特点%Features of Reactor Structure Design for 10MW High Temperature Gas-Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    刘俊杰; 王敏稚; 张征明; 张振声; 何树延

    2001-01-01

    模块式高温气冷堆是当今世界上公认的先进反应堆堆型之一。固有安全性是它的最突出的优点。本文对 10 MW高温气冷堆的堆体布置进行了详细描述,并对 10MW高温气冷堆的结构设计特点进行了分析。根据 10 MW高温气冷堆的特点,本文对该堆的固有安全性、制造工艺等方面的优点进行了论述。%As the world-wide accepted advanced nuclear reactor,the most prominent characteristic of High-temperature Gas-cooled Modular Reactor is its inherent safety.The core arrangement of the 10MW High-temperature Gas-cooled Reactor,which is now under construction,is described in detail.The features of its structure design are analyzed,and based on these features,the advantages of inherent safety and manufacture are also discussed.

  3. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  4. Simulation of operational an accidental behaviour of modular high temperature reactors with Brayton cycle Power Conversion Unit

    International Nuclear Information System (INIS)

    The present work analyses and investigates the behaviour of a High Temperature Reactor (HTR) with a Pebble Bed core connected to a Brayton cycle Power Conversion Unit (PCU) during operational and accident conditions. The modelling of a complete circuit including both the PCU and the Pebble Bed Reactor has been performed with the commercial thermal-fluid analysis simulation code Flownex. Flownex has been developed for High Temperature Pebble Bed Reactor applications, and has been exten-sively validated against other codes. As the reactor core model incorporated in Flownex is a simplified model based on 0D point kinetics, the extended 1D WKIND core model was implemented in the analysis calculations using a special coupling methodology. This study introduces a new sub-routine which enables the cou-pling of the WKIND reactor core model to the Flownex PCU model via an external interface. The interface facilitates the data exchange between the two codes, allowing for necessary manipulations and synchronisation of the coupled codes. By doing so, the 1D diffusion equation solution implemented in WKIND core model replaces the point kinetics model implemented in Flownex. This replacement allows for a detailed accurate solution even for very fast transients, through the treatment of the space-dependent heat conduction from the graphite matrix to helium. Flownex component models have been validated against the experimental results of the 50 MWel direct helium turbine facility Energieversorgung Oberhausen (EVO II). This provided the opportunity to validate Flownex calculations against experimental data derived from a large-scale helium Brayton cycle installation. Small differences observed in the results could be explained. Based upon steady state and transient analysis it is concluded that Flownex models simulate accurately the behaviour of the components integrated in the EVO II plant. Such models could be applied to analyse the transient behaviour of the total system of the

  5. Shutdown cooling helium circulator design considerations for MHTGR [Modular High Temperature Gas-Cooled Reactor] power plant

    International Nuclear Information System (INIS)

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) plant embodies a shutdown cooling system to expedite plant cooldown for refueling, maintenance, and repair in the event that the main cooling loop is unavailable. This is a non safety related system. A key component in this system, is a helium circulator. Oriented vertically, the rotating assembly in this machine is supported on active magnetic bearings, and the radial flow compressor is driven by a submerged induction electric motor rated at 160 kW(e). This paper gives details of the circulator design considerations and includes topics related to the machine operation and maintenance, and the technology base. 12 refs., 11 figs., 3 tabs

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

    International Nuclear Information System (INIS)

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

  7. Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

    International Nuclear Information System (INIS)

    The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

  8. ICARUS.2: a kinetics analysis code for the experimental very high temperature gas-cooled reactor plant

    International Nuclear Information System (INIS)

    The report describes analytical models, computational methods and physical properties, which are employed in the computational code ICARUS.2. The code simulates the transient behavior of the experimental Very High Temperature Reactor plant. Transient responses of the plant to various disturbances are analyzed as well as of the typical components in the plant. The code has the enough number of modeled components to cover the two-loop system of the real plant. Each loop consists of three internal coolant loops which are the primary and the secondary helium coolant loops and the ternary water-steam coolant loop. Further, the code can model any structure of plant coolant system which might cause the significant effect on the transient behavior. (author)

  9. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  10. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

  11. Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; YANG Rui-Chang; JIA Dou-Nan

    2005-01-01

    The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calculated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen's formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell's temperature. Its maximum value is 14%.

  12. Assessment of fuel integrity of the High Temperature Engineering Test Reactor (HTTR) and its permissible design limit

    International Nuclear Information System (INIS)

    This report describes the results of integrity assessment of the HTTR fuel as a part of safety design of the reactor. Functions of individual coating layer of the coated fuel particles were investigated to clarify the characteristic properties of the HTTR fuel. Integrity of the fuel under normal operating conditions of HTTR was shown to be maintained through analyses on fuel failure and due to kernel migration (amoeba effect) and corrosion of the silicon carbide layer by palladium. The fuel integrity has been demonstrated by successful results of the irradiation performance tests. By considering the failure behavior of coated fuel particles, the permissible design limit of the fuel was determined such that the fuel temperature should not exceed 1600degC under abnormal transient conditions. The validity of the determination of the limit has been confirmed by results of heating experiments of coated fuel particles at extremely high temperature. (author)

  13. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (2)

    International Nuclear Information System (INIS)

    This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)

  14. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

  15. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  16. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  17. Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

    International Nuclear Information System (INIS)

    The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 deg. C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 deg. C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 deg. C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating

  18. High-temperature behavior of dicesium molybdate Cs2MoO4: Implications for fast neutron reactors

    Science.gov (United States)

    Wallez, Gilles; Raison, Philippe E.; Smith, Anna L.; Clavier, Nicolas; Dacheux, Nicolas

    2014-07-01

    Dicesium molybdate (Cs2MoO4)'s thermal expansion and crystal structure have been investigated herein by high temperature X ray diffraction in conjunction with Raman spectroscopy. This first crystal-chemical insight at high temperature is aimed at predicting the thermostructural and thermomechanical behavior of this oxide formed by the accumulation of Cs and Mo fission products at the periphery of nuclear fuel rods in sodium-cooled fast reactors. Within the temperature range of the fuel's rim, Cs2MoO4 becomes hexagonal P63/mmc, with disordered MoO4 tetrahedra and 2D distribution of Cs-O bonds that makes thermal axial expansion both large (50≤αl≤70 10-6 °C-1, 500-800 °C) and highly anisotropic (αc-αa=67×10-6 °C-1, hexagonal form). The difference with the fuel's expansion coefficient is of potential concern with respect to the cohesion of the Cs2MoO4 surface film and the possible release of cesium radionuclides in accidental situations.

  19. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    ) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of

  20. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors 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 United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. 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, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  1. Processing of mixed uranium/refractory metal carbide fuels for high temperature space nuclear reactors

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim

    2000-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for high performance, next generation space power and propulsion systems. These mixed carbides such as the pseudo-ternary, (U, Zr, Nb)C, hold significant promise because of their high melting points (typically greater than 3200 K), thermochemical stability in a hot hydrogen environment, and high thermal conductivity. However, insufficient test data exist under nuclear thermal propulsion conditions of temperature and hot hydrogen environment to fully evaluate their performance. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders while hypostoichiometric samples with carbon-to-metal (C/M) ratios of 0.95 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold pressing, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce dense (low porosity), homogeneous, single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for space power and propulsion applications. .

  2. Advanced processing of lead titanate-polyimide composites for high temperature piezoelectric sensing

    NARCIS (Netherlands)

    Khanbareh, H.; Hegde, M.; Zwaag, S. van der; Groen, W.A.

    2015-01-01

    High performance polymer-ceramic composites are presented as promising candidates for high temperature piezoelectric sensing applications. lead-titanate (PT) ceramic particulate is incorporated into a polyetherimide polymer matrix, (PEI) at a specific volume fraction of 20% in the forms of 0-3 and q

  3. Review on an Advanced High-Temperature Measurement Technology: The Optical Fiber Thermometry

    Directory of Open Access Journals (Sweden)

    Y. B. Yu

    2009-01-01

    Full Text Available Optical fiber thermometry technology for high-temperature measurement is briefly reviewed in this paper. The principles, characteristics, recent progresses and advantages of the technology are described. Examples of using the technology are introduced. Many blackbody, infrared, and fluorescence optical thermometers are developed for practical applications.

  4. Theoretical and experimental research of natural convection in the core of the high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The physical model of the developed THERMIX-2D-code for computing thermohydraulic behaviour of the core of high temperature pebble bed reactors is verified by experiments with natural convection flow. Such fluid flow behaviour can be of very high importance for the real reactor in the case of natural heat removal decay. The experiments are performed in a special set up testing-stand with pressures up to 30 bars and temperatures up to 3000C by using air and helium as fluid. In comparison with the experimental data the numerical results show that a good and useful simulation is given by the program. Pure natural convection flow in packed pebble beds is calculated with a very high degree of reliability. The investigation of flow stability demonstrate that radial-symmetric relations are not given temporarily when national convection is overlayed by forced convection flow. In the discussion it is explained when and to what extent the program leds to useful results in such situations. The test of the effective heat conductivity lambdasub(eff) results in an improvement of the lambdasub(eff)-data used so far for temperatures below 13000C. (orig.)

  5. Papers concerning fuel particles, fuel elements and graphitic materials for high temperature reactors presented at the 1980 annual conference Kerntechnik

    International Nuclear Information System (INIS)

    This report is a compilation of the papers presented by staff of the Institute for Reactor Materials, KFA-Juelich, at the 1980 annual conference Rekatortechnik, held in Berlin, 25-27th March 1980. In some cases, there were co-authors from other organisations. Where possible the manuscripts of the presentations have been reproduced, as well as the display cards shown during the poster session on the conference. In the presentations, the questions of the characterization of fuel particles and the retention of fission products are dealt with, and special attention is given to fission product release at very high temperatures. One presentation deals with the disposal of fuel elements from the AVR. Another report presents the results of radiation experiments on the standard matrix material A3-3 of the THTR fuel elements; the changes in dimensions, creep coefficient and thermal conductivity were measured as functions of the fluence and the radiation temperature. Interim results obtained from long-term radiation experiments on reflector graphites for high and very high flux reactors are presented, and models for the calculation of dimensional changes in components subjected to fluctuating temperatures from data obtained in isothermal tests are discussed. (orig.)

  6. Using high temperature gas-cooled reactors for greenhouse gas reduction and energy neutral production of phosphate fertilizers

    International Nuclear Information System (INIS)

    Highlights: • We estimate the energy requirements of wet- and thermal phosphate rock processing. • We estimate the amount of U needed to operate a representative HTGR. • Energy neutral phosphate fertilizer production is theoretically possible. - Abstract: This paper discusses how high temperature gas-cooled reactors (HTGRs) could provide energy for phosphate rock (PR) processing while extracting uranium (U) from the processed PR that can again be used as raw material for nuclear reactor fuel that may power the greenhouse gas lean energy source employed. First estimates using a HTGR presently constructed in China (HTR-PM) conclude that a concentration of approximately 80 mg/kg U in PR is sufficiently high for energy neutral wet acid PR processing with waste treatment and a concentration of approximately 110 mg/kg U is adequate to promote energy intensive high quality thermal phosphoric acid production. In addition, the recovery of U from PRs yields beneficial side-effects in a way that U loads on agricultural soils are reduced and consequently contamination of groundwater with U will be diminished

  7. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  8. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  9. Gigawatt-year nuclear-geothermal energy storage for light-water and high-temperature reactors

    International Nuclear Information System (INIS)

    Capital-intensive, low-operating cost nuclear plants are most economical when operated under base-load conditions. However, electricity demand varies on a daily, weekly, and seasonal basis. In deregulated utility markets this implies high prices for electricity at times of high electricity demand and low prices for electricity at times of low electricity demand. We examined coupling nuclear heat sources to geothermal heat storage systems to enable these power sources to meet hourly to seasonal variable electricity demand. At times of low electricity demand the reactor heats a fluid that is then injected a kilometer or more underground to heat rock to high temperatures. The fluid travels through the permeable-rock heat-storage zone, transfers heat to the rock, is returned to the surface to be reheated, and re-injected underground. At times of high electricity demand the cycle is reversed, heat is extracted, and the heat is used to power a geothermal power plant to produce intermediate or peak power. When coupling geothermal heat storage with light-water reactors (LWRs), pressurized water (<300 deg. C) is the preferred heat transfer fluid. When coupling geothermal heat storage with high temperature reactors at higher temperatures, supercritical carbon dioxide is the preferred heat transfer fluid. The non-ideal characteristics of supercritical carbon dioxide create the potential for efficient coupling with supercritical carbon dioxide power cycles. Underground rock cannot be insulated, thus small heat storage systems with high surface to volume ratios are not feasible because of excessive heat losses. The minimum heat storage capacity to enable seasonal storage is ∼0.1 Gigawatt-year. Three technologies can create the required permeable rock: (1) hydro-fracture, (2) cave-block mining, and (3) selective rock dissolution. The economic assessments indicated a potentially competitive system for production of intermediate load electricity. The basis for a nuclear

  10. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  11. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  12. High temperature ceramic membrane reactors for coal liquid upgrading. Final report, September 21, 1989--November 20, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T. [University of Southern California, Los Angeles, CA (United States). Dept. of Chemical Engineering; Liu, P.K.T. [Aluminum Co. of America, Pittsburgh, PA (United States); Webster, I.A. [Unocal Corp., Los Angeles, CA (United States)

    1992-12-31

    Membrane reactors are today finding extensive applications for gas and vapor phase catalytic reactions (see discussion in the introduction and recent reviews by Armor [92], Hsieh [93] and Tsotsis et al. [941]). There have not been any published reports, however, of their use in high pressure and temperature liquid-phase applications. The idea to apply membrane reactor technology to coal liquid upgrading has resulted from a series of experimental investigations by our group of petroleum and coal asphaltene transport through model membranes. Coal liquids contain polycyclic aromatic compounds, which not only present potential difficulties in upgrading, storage and coprocessing, but are also bioactive. Direct coal liquefaction is perceived today as a two-stage process, which involves a first stage of thermal (or catalytic) dissolution of coal, followed by a second stage, in which the resulting products of the first stage are catalytically upgraded. Even in the presence of hydrogen, the oil products of the second stage are thought to equilibrate with the heavier (asphaltenic and preasphaltenic) components found in the feedstream. The possibility exists for this smaller molecular fraction to recondense with the unreacted heavy components and form even heavier undesirable components like char and coke. One way to diminish these regressive reactions is to selectively remove these smaller molecular weight fractions once they are formed and prior to recondensation. This can, at least in principle, be accomplished through the use of high temperature membrane reactors, using ceramic membranes which are permselective for the desired products of the coal liquid upgrading process. An additional incentive to do so is in order to eliminate the further hydrogenation and hydrocracking of liquid products to undesirable light gases.

  13. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was 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 CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  14. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    International Nuclear Information System (INIS)

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was 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. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a variety

  15. Advances in processing of NiAl intermetallic alloys and composites for high temperature aerospace applications

    Science.gov (United States)

    Bochenek, Kamil; Basista, Michal

    2015-11-01

    Over the last few decades intermetallic compounds such as NiAl have been considered as potential high temperature structural materials for aerospace industry. A large number of investigations have been reported describing complex fabrication routes, introducing various reinforcing/alloying elements along with theoretical analyses. These research works were mainly focused on the overcoming of main disadvantage of nickel aluminides that still restricts their application range, i.e. brittleness at room temperature. In this paper we present an overview of research on NiAl processing and indicate methods that are promising in solving the low fracture toughness issue at room temperature. Other material properties relevant for high temperature applications are also addressed. The analysis is primarily done from the perspective of NiAl application in aero engines in temperature regimes from room up to the operating temperature (over 1150 °C) of turbine blades.

  16. High temperature on-line monitoring of water chemistry and corrosion control in water cooled power reactors. Report of a co-ordinated research project 1995-1999

    International Nuclear Information System (INIS)

    This report documents the results of the Co-ordinated Research Project (CRP) on High Temperature On-line Monitoring of Water Chemistry and Corrosion in Water Cooled Power Reactors (1995-1999). This report attempts to provide both an overview of the state of the art with regard to on-line monitoring of water chemistry and corrosion in operating reactors, and technical details of the important contributions made by programme participants to the development and qualification of new monitoring techniques. The WACOL CRP is a follow-up to the WACOLIN (Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors) CRP conducted by the IAEA from 1986 to 1991. The WACOLIN CRP, which described chemistry, corrosion and activity-transport aspects, clearly showed the influence of water chemistry on corrosion of both fuel and reactor primary-circuit components, as well as on radiation fields. It was concluded that there was a fundamental need to monitor water-chemistry parameters in real time, reliably and accurately. The objectives of the WACOL CRP were to establish recommendations for the development, qualification and plant implementation of methods and equipment for on-line monitoring of water chemistry and corrosion. Chief investigators from 18 organizations representing 15 countries provided a variety of contributions aimed at introducing proven monitoring techniques into plants on a regular basis and filling the gaps between plant operator needs and available monitoring techniques. The CRP firmly demonstrated that in situ monitoring is able to provide additional and valuable information to plant operators, e.g. ECP, high temperature pH and conductivity. Such data can be obtained promptly, i.e. in real time and with a high degree of accuracy. Reliable techniques and sensor devices are available which enable plant operators to obtain additional information on the response of structural materials in

  17. AMSAHTS 1990: Advances in Materials Science and Applications of High-Temperature Superconductors

    International Nuclear Information System (INIS)

    This publication is comprised of abstracts for oral and poster presentations scheduled for AMSAHTS '90. The conference focused on understanding high temperature superconductivity with special emphasis on materials issues and applications. AMSAHTS 90, highlighted the state of the art in fundamental understanding of the nature of high-Tc superconductivity (HTSC) as well as the chemistry, structure, properties, processing and stability of HTSC oxides. As a special feature of the conference, space applications of HTSC were discussed by NASA and Navy specialists

  18. Steam oxidation of advanced high temperature resistant alloys for ultra-supercritical applications

    OpenAIRE

    Lukaszewicz, Mikolaj

    2012-01-01

    Steam oxidation of heat exchanger tubing is of growing interest as increasing the efficiencies of conventional pulverised fuel fired power plants requires higher steam temperatures and pressures. These new, more severe steam conditions result in faster steam oxidation reactions, which can significantly reduce the lifetime of boiler components. This thesis reports results from an investigation of the steam oxidation of the high temperature resistant alloys. It covers an analysis of the impact ...

  19. A National Collaboratory to Advance the Science of High Temperature Plasma Physics for Magnetic Fusion

    International Nuclear Information System (INIS)

    This report summarizes the work of the National Fusion Collaboratory (NFC) Project funded by the United States Department of Energy (DOE) under the Scientific Discovery through Advanced Computing Program (SciDAC) to develop a persistent infrastructure to enable scientific collaboration for magnetic fusion research. A five year project that was initiated in 2001, it built on the past collaborative work performed within the U.S. fusion community and added the component of computer science research done with the USDOE Office of Science, Office of Advanced Scientific Computer Research. The project was a collaboration itself uniting fusion scientists from General Atomics, MIT, and PPPL and computer scientists from ANL, LBNL, Princeton University, and the University of Utah to form a coordinated team. The group leveraged existing computer science technology where possible and extended or created new capabilities where required. Developing a reliable energy system that is economically and environmentally sustainable is the long-term goal of Fusion Energy Science (FES) research. In the U.S., FES experimental research is centered at three large facilities with a replacement value of over $1B. As these experiments have increased in size and complexity, there has been a concurrent growth in the number and importance of collaborations among large groups at the experimental sites and smaller groups located nationwide. Teaming with the experimental community is a theoretical and simulation community whose efforts range from applied analysis of experimental data to fundamental theory (e.g., realistic nonlinear 3D plasma models) that run on massively parallel computers. Looking toward the future, the large-scale experiments needed for FES research are staffed by correspondingly large, globally dispersed teams. The fusion program will be increasingly oriented toward the International Thermonuclear Experimental Reactor (ITER) where even now, a decade before operation begins, a large

  20. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

    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 Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  1. Additive Manufacturing of Advanced High Temperature Masking Fixtures for EBPVD TBC Coating

    Energy Technology Data Exchange (ETDEWEB)

    List, III, Frederick Alyious [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feuerstein, Albert [Praxair Surface Technologies, Inc., (United States); Dehoff, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kirka, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carver, Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-03-30

    The purpose of this Manufacturing Demonstration Facility (MDF) technical collaboration project between Praxair Surface Technologies, Inc. (PST) and Oak Ridge National Laboratory (ORNL) was to develop an additive manufacturing process to fabricate next generation high temperature masking fixtures for coating of turbine airfoils with ceramic Thermal Barrier Coatings (TBC) by the Electron Beam Physical Vapor Deposition (EBPVD) process. Typical masking fixtures are sophisticated designs and require complex part manipulation in order to achieve the desired coating distribution. Fixtures are typically fabricated from high temperature nickel (Ni) based superalloys. The fixtures are fabricated from conventional processes by welding of thin sheet material into a complex geometry, to decrease the weight load for the manipulator and to reduce the thermal mass of the fixture. Recent attempts have been made in order to fabricate the fixtures through casting, but thin walled sections are difficult to cast and have high scrap rates. This project focused on understanding the potential for fabricating high temperature Ni based superalloy fixtures through additive manufacturing. Two different deposition processes; electron beam melting (EBM) and laser powder bed fusion were evaluated to determine the ideal processing route of these materials. Two different high temperature materials were evaluated. The high temperature materials evaluated were Inconel 718 and another Ni base alloy, designated throughout the remainder of this document as Alloy X, as the alloy composition is sensitive. Inconel 718 is a more widely utilized material for additive manufacturing although it is not currently the material utilized for current fixtures. Alloy X is the alloy currently used for the fixtures, but is not a commercially available alloy for additive manufacturing. Praxair determined it was possible to build the fixture using laser powder bed technology from Inconel 718. ORNL fabricated the fixture

  2. Advanced Packaging Technology Used in Fabricating a High-Temperature Silicon Carbide Pressure Sensor

    Science.gov (United States)

    Beheim, Glenn M.

    2003-01-01

    The development of new aircraft engines requires the measurement of pressures in hot areas such as the combustor and the final stages of the compressor. The needs of the aircraft engine industry are not fully met by commercially available high-temperature pressure sensors, which are fabricated using silicon. Kulite Semiconductor Products and the NASA Glenn Research Center have been working together to develop silicon carbide (SiC) pressure sensors for use at high temperatures. At temperatures above 850 F, silicon begins to lose its nearly ideal elastic properties, so the output of a silicon pressure sensor will drift. SiC, however, maintains its nearly ideal mechanical properties to extremely high temperatures. Given a suitable sensor material, a key to the development of a practical high-temperature pressure sensor is the package. A SiC pressure sensor capable of operating at 930 F was fabricated using a newly developed package. The durability of this sensor was demonstrated in an on-engine test. The SiC pressure sensor uses a SiC diaphragm, which is fabricated using deep reactive ion etching. SiC strain gauges on the surface of the diaphragm sense the pressure difference across the diaphragm. Conventionally, the SiC chip is mounted to the package with the strain gauges outward, which exposes the sensitive metal contacts on the chip to the hostile measurement environment. In the new Kulite leadless package, the SiC chip is flipped over so that the metal contacts are protected from oxidation by a hermetic seal around the perimeter of the chip. In the leadless package, a conductive glass provides the electrical connection between the pins of the package and the chip, which eliminates the fragile gold wires used previously. The durability of the leadless SiC pressure sensor was demonstrated when two 930 F sensors were tested in the combustor of a Pratt & Whitney PW4000 series engine. Since the gas temperatures in these locations reach 1200 to 1300 F, the sensors were

  3. Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

    International Nuclear Information System (INIS)

    Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

  4. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  5. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  6. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

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

  7. Advanced high temperature materials for the energy efficient automotive Stirling engine

    International Nuclear Information System (INIS)

    The Stirling engine is under investigation jointly by the Department of Energy and NASA Lewis as an alternative to the internal combustion engine for automotive applications. The Stirling engine is an external combustion engine that offers the advantage of high fuel economy, low emissions, low noise, and low vibrations compared to current internal combustion automotive engines. The most critical component from a materials viewpoint is the heater head consisting of the cylinders, heating tubes, and regenerator housing. Materials requirements for the heater head include compatibility with hydrogen, resistance to hydrogen permeation, high temperature oxidation/corrosion resistance, and high temperature creep-rupture and fatigue properties. A continuing supporting materials research and technology program has identified the wrought alloys CG-27 and 12RN72, and the cast alloys XF-818 and NASAUT 4G-A1 as candidate replacements for the cobalt containing alloys used in current prototype engines. Based on the materials research program in support of the automotive Stirling engine it is concluded that manufacture of the engine is feasible from low cost iron-base alloys rather than the cobalt alloys used in prototype engines. This paper presents results of research that led to this conclusion

  8. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was 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 CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  9. Analytical estimation of control rod shadowing effect for excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki; Yamashita, Kiyonobu; Fujimoto, Nozomu; Nojiri, Naoki; Takeuchi, Mitsuo; Fujisaki, Shingo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tokuhara, Kazumi; Nakata, Tetsuo

    1998-05-01

    The control rod shadowing effect has been estimated analytically in application of the fuel addition method to excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR). The movements of control rods in the procedure of the fuel addition method have been simulated in the analysis. The calculated excess reactivity obtained by the simulation depends on the combinations of measuring control rods and compensating control rods and varies from -10% to +50% in comparison with the excess reactivity calculated from the effective multiplication factor of the core where all control rods are fully withdrawn. The control rod shadowing effect is reduced by the use of plural number of measuring and compensation control rods because of the reduction in neutron flux deformation in the measuring procedure. As a result, following combinations of control rods are recommended; 1) Thirteen control rods of the center, first, and second rings will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other twelve control rods for reactivity compensation. 2) Six control rods of the first ring will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other five control rods for reactivity compensation. (author)

  10. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.