FBR and RBR particle bed space reactors

International Nuclear Information System (INIS)

Powell, J.R.; Botts, T.E.

1983-01-01

Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10 0 K), high coolant-outlet temperatures (1500 to 3000 0 K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

International Nuclear Information System (INIS)

Gokhan Yesilyurt; Hassan, Y.A.

2003-01-01

A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

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

International Nuclear Information System (INIS)

Schuerenkraemer, M.

1984-04-01

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 300 0 C 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 1300 0 C. (orig.) [de

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

International Nuclear Information System (INIS)

Zheng Yanhua; Shi Lei; Wang Yan

2010-01-01

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

Modular high-temperature reactor launched (and wallchart)

International Nuclear Information System (INIS)

Steinwarz, W.

1987-01-01

In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

Porous structure analysis of large-scale randomly packed pebble bed in high temperature gas-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Ren, Cheng; Yang, Xingtuan; Liu, Zhiyong; Sun, Yanfei; Jiang, Shengyao [Tsinghua Univ., Beijing (China). Key Laboratory of Advanced Reactor Engineering and Safety; Li, Congxin [Ministry of Environmental Protection of the People' s Republic of China, Beijing (China). Nuclear and Radiation Safety Center

2015-02-15

A three-dimensional pebble bed corresponding to the randomly packed bed in the heat transfer test facility built for the High Temperature Reactor Pebble bed Modules (HTR-PM) in Shandong Shidaowan is simulated via discrete element method. Based on the simulation, we make a detailed analysis on the packing structure of the pebble bed from several aspects, such as transverse section image, longitudinal section image, radial and axial porosity distributions, two-dimensional porosity distribution and coordination number distribution. The calculation results show that radial distribution of porosity is uniform in the center and oscillates near the wall; axial distribution of porosity oscillates near the bottom and linearly varies along height due to effect of gravity; the average coordination number is about seven and equals to the maximum coordination number frequency. The fully established three-dimensional packing structure analysis of the pebble bed in this work is of fundamental significance to understand the flow and heat transfer characteristics throughout the pebble-bed type structure.

Fluidized-bed nuclear reactor

International Nuclear Information System (INIS)

Grimmett, E.S.; Kunze, J.F.

1975-01-01

A reactor vessel containing a fluidized-bed region of particulate material including both a neutron-moderating and a fertile substance is described. A gas flow including fissile material passes through the vessel at a sufficient rate to fluidize the particulate material and at a sufficient density to support a thermal fission reaction within the fluidized-bed region. The high-temperature portion of a heat transfer system is located within the fluidized-bed region of the reactor vessel in direct contact with the fluidized particles. Heat released by fission is thereby transferred at an enhanced rate to a coolant circulating within the heat transfer system. Fission products are continuously removed from the gas flow and supplemental fissile material added during the reactor operation. (U.S.)

The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

2012-01-01

The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Sen, R. S.; Cogliati, J. J.; Gougar, H. D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2012-07-01

The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

International Nuclear Information System (INIS)

Nabbi, R.; Meister, G.; Finken, R.; Haben, M.

1982-09-01

The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

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.

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

International Nuclear Information System (INIS)

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

1981-01-01

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

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

International Nuclear Information System (INIS)

Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

1987-01-01

Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

Numerical investigation of the 3-dimensional steady-state temperature- and flow distribution in the core of a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Verfondern, K.

1983-01-01

This work presents a computer model determining the steady-state temperature- and flow field in 3 dimensions in the core of a pebble bed high temperature reactor. The numerical sprinkler method, basind on the Thermix-model, allows to describe the thermo-hydraulics of a non-rotational-symmetric core-geometry. The AVR-reactor in Juelich, in operation since 1967, represents a suitable investigation-object for the computer model of Thermix-3D. It is in a 3D-mesh-structure to reproduce very precisely the so called ''graphite noses'', in which the shut-down rods are conducted as well as the filling cones in the inner and outer area. The results of the final calculation of the normal operation condition for the AVR-reactor unambiguously show, that within the core reproduced in 3 dimensions there are evident deviations in the flow profile and in the temperatures of the cooling gas in contrast to a 2D-handling. (orig.) [de

Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

Directory of Open Access Journals (Sweden)

Shixiong Song

2014-01-01

CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

Development of high temperature gas cooled reactor in China

Energy Technology Data Exchange (ETDEWEB)

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

2018-02-15

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

Numerical investigation of the flow at the pebble bed of the high temperature gas cooled reactors

International Nuclear Information System (INIS)

Costa, Franklin C.; Navarro, Moyses A.; Santos, Andre A.C.

2011-01-01

This paper presents a numerical investigation of the thermal and fluid dynamics among the fuel spheres and the cooling fluid, appearing in the core of pebble bed reactor (PBR-Peeble Bed Reactor) using the CFD-Computational Fluid Dynamics CFX 13.0. The paper presents the two analysis results. In the first phase it was considered two heat transfer models for the fuel spheres. In a model it was established volumetric load generation, with thermal conduction for both the fuel and coating. The other model prescribes a heat flux at the sphere surfaces. In this analysis, it was proceed two simulation in the two sphere arrangements, one considering the spheres in contact, and the other with 2 mm spacing between them. At the second analysis it was evaluated the sphere arrangement influence on the thermal and fluid dynamic behavior of the bed. The four simulations present differences in the flow and in the surface and maximum temperature profiles of the coating.(author)

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

Science.gov (United States)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

International Nuclear Information System (INIS)

Zhu, G.; Zou, Y.; Xu, H.

2016-01-01

Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PBFHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF_2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-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 a good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel. (A.C)

Computational and experimental prediction of dust production in pebble bed reactors, Part II

Energy Technology Data Exchange (ETDEWEB)

Hiruta, Mie; Johnson, Gannon [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Rostamian, Maziar, E-mail: mrostamian@asme.org [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Potirniche, Gabriel P. [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Ougouag, Abderrafi M. [Idaho National Laboratory, 2525 N Fremont Avenue, Idaho Falls, ID 83401 (United States); Bertino, Massimo; Franzel, Louis [Department of Physics, Virginia Commonwealth University, Richmond, VA 23284 (United States); Tokuhiro, Akira [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States)

2013-10-15

Highlights: • Custom-built high temperature, high pressure tribometer is designed. • Two different wear phenomena at high temperatures are observed. • Experimental wear results for graphite are presented. • The graphite wear dust production in a typical Pebble Bed Reactor is predicted. -- Abstract: This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

Chemical-looping combustion in a reverse-flow fixed bed reactor

International Nuclear Information System (INIS)

Han, Lu; Bollas, George M.

2016-01-01

A reverse-flow fixed bed reactor concept for CLC (chemical-looping combustion) is explored. The limitations of conventional fixed bed reactors, as applied to CLC, are overcome by reversing the gas flow direction periodically to enhance the mixing characteristics of the bed, thus improving oxygen carrier utilization and energy efficiency with respect to power generation. The reverse-flow reactor is simulated by a dusty-gas model and compared with an equivalent fixed bed reactor without flow reversal. Dynamic optimization is used to calculate conditions at which each reactor operates at maximum energy efficiency. Several cases studies illustrate the benefits of reverse-flow operation for the CLC with CuO and NiO oxygen carriers and methane and syngas fuels. The results show that periodic reversal of the flow during reduction improves the contact between the fuel and unconverted oxygen carrier, enabling the system to suppress unwanted catalytic reactions and axial temperature and conversion gradients. The operational scheme presented reduces the fluctuations of temperature during oxidation and increases the high-temperature heat produced by the process. CLC in a reverse-flow reactor has the potential to achieve higher energy efficiency than conventional fixed bed CLC reactors, when integrated with a downstream gas turbine of a combined cycle power plant. - Highlights: • Reverse-flow fixed bed CLC reactors for combined cycle power systems. • Dynamic optimization tunes operation of batch and transient CLC systems. • The reverse-flow CLC system provides stable turbine-ready gas stream. • Reverse-flow CLC fixed bed reactor has superior CO 2 capture and thermal efficiency.

Researchers solve big mysteries of pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Shams, Afaque; Roelofs, Ferry; Komen, E.M.J. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Baglietto, Emilio [Massachusetts Institute of Technology, Cambridge, MA (United States). Dept. of Nuclear Science and Engineering; Sgro, Titus [CD-adapco, London (United Kingdom). Technical Marketing

2014-03-15

The PBR is one type of High Temperature Reactors, which allows high temperature work while preventing the fuel from melting (bringing huge safety margins to the reactor) and high electricity efficiency. The design is also highly scalable; a plant could be designed to be as large or small as needed, and can even be made mobile, allowing it to be used onboard a ship. In a PBR, small particles of nuclear fuel, embedded in a moderating graphite pebble, are dropped into the reactor as needed. At the bottom, the pebbles can be removed simply by opening a small hatch and letting gravity pull them down. To cool the reactor and create electricity, helium gas is pumped through the reactor to pull heat out which is then run through generators. One of the most difficult problems to deal with has been the possible appearance of local temperature hotspots within the pebble bed heating to the point of melting the graphite moderators surrounding the fuel. Obviously, constructing a reactor and experimenting to investigate this possibility is out of the question. Instead, nuclear engineers have been attempting to simulate a PBR with various CFD codes. The thermo-dynamic analysis to simulate realistic conditions in a pebble bed are described and the results are shown. (orig.)

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

Energy Technology Data Exchange (ETDEWEB)

Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

2017-07-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

International Nuclear Information System (INIS)

Moreira, Uebert G.; Dominguez, Dany S.

2017-01-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

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...... into long cylindrical pellets. Kaolin and bituminous coal ash that both have significant amounts of Si and Al show superior potassium capture characteristics. Experimental results show that capture of potassium by kaolin is independent of the gas oxygen content. Kaolin releases water and forms metakaolin...... when heated at temperatures above 450°C. The amounts of potassium captured by metakaolin pellet decreases with increasing reaction temperature in the range of 900-1300°C and increases again with further increasing the temperature up to 1500°C. There is no reaction of pre-made mullite with KCl...

Neutron physical investigations on the shutdown effect of small boronated absorbing spheres for pebble-bed high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Sgouridis, S.; Schurrer, F.; Muller, H.; Ninaus, W.; Oswald, K.; Neef, R.D.; Schaal, H.

1987-01-01

An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres

Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled High Temperature Reactor - 15171

International Nuclear Information System (INIS)

Zhu, G.; Zou, Y.; Xu, Hongjie

2015-01-01

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-BeF 2 ) salt temperature reactivity coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing heavy metal loading and decreasing excessive moderation. In order to analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared 2 refueling schemes (mixing flow pattern and directional flow pattern) and 2 kinds of reflector materials (SiC and graphite). This method has found that the feasible regions of breeding and negative FLiBe TRC is between 20 vol% and 62 vol% heavy metal 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 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margins. The greatest challenge of this reactor may be the very long irradiation time of the pebble fuel. (authors)

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

International Nuclear Information System (INIS)

2002-01-01

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

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

International Nuclear Information System (INIS)

Bjornard, Trond; Hockert, John

2011-01-01

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

High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

International Nuclear Information System (INIS)

Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

1993-01-01

An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

High temperature fusion reactor design

International Nuclear Information System (INIS)

Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

1979-01-01

Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

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

International Nuclear Information System (INIS)

Teuchert, E.; Ruetten, H.J.; Worley, B.A.; Vondy, D.R.

1979-11-01

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 233 U inventory for a breeder could be produced by prebreeders using 235 U fuel

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

International Nuclear Information System (INIS)

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

2005-01-01

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

High-temperature reactor developments in the Netherlands

International Nuclear Information System (INIS)

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.)

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.

Experimental investigation on feasibility of two-region-designed pebble-bed high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Yang Xingtuan; Hu Wenping; Jiang Shengyao

2009-01-01

Phenomenological experiments were performed on a 2-dimensional scaled model of the two-region designed pebble-bed high-temperature gas-cooled reactor core consisting of the distinct fuel pebble region and graphite pebble region. Issues with respect to the feasibility of the two-region design, including the establishment of the two-region arrangement, the mixing zone between the two regions, and the stagnant zone existence, were investigated. Three equilibrium conditions were proposed to evaluate the stable two-region arrangement formation. The general characteristics of the flow of the pebble bed were analyzed on basis of the observed phenomenon. It was found that a stable two-region arrangement was formed under the experimental conditions: the pebbles' motion was to some extent random but also confined by the neighbors of pebbles so that the mixing zone is constrained to a reasonable size. Guide plates utilized to improve mixing are proved to be effective without noticeable effect on the two-region arrangement features. Stagnant zones were observed under the experimental conditions and they were expected to be avoided by improving the design of the experimental setup. (author)

The importance of the AVR pebble-bed reactor for the future of nuclear power

International Nuclear Information System (INIS)

Pohl, P.

2006-01-01

The AVR pebble-bed high temperature gas-cooled reactor (HTGR) at Juelich (Germany)) operated from 1967 to 1988 and was certainly the most important HTGR project of the past. The reactor was the mass test bed for all development steps of HTGR pebble fuel. Some early fuel charges failed under high temperature conditions and contaminated the reactor. An accurate pebble measurement (Cs 137) allowed to clean the core from unwanted pebbles after 1981. The coolant activity went down and remained very low for the remaining reactor operation. A melt-wire experiment in 1986 revealed max. coolant temperatures of >1280 deg. C and fuel temperatures of >1350 deg. C, explained by under-estimated bypasses. The fuel still in the core achieved high burn-ups and showed under the extreme temperature conditions excellent fission product retention. Thus, the AVR operation qualified the HTGR fuel, and an average discharge burn-up of 112% fifa revealed an excellent fuel economy of the pebble-bed reactor. Furthermore, the AVR operation offers many meaningful data for code-to-experiment comparisons. (authors)

Parametric study for high conversion pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Teuchert, E.; Ruetten, H. J.

1975-06-15

Tables are presented of fuel cycle costs, conversion ratios and accompanying variations in fuel element designs for a 3,00 MWth high conversion pebble bed reactor with initial high enriched uranium/thorium cycle and subsequent recycling of U-233, Pu-239 and Pu-241.

A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

2016-07-15

Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

New deployment of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

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

2002-01-01

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

Computational and experimental prediction of dust production in pebble bed reactors, Part II

Energy Technology Data Exchange (ETDEWEB)

Mie Hiruta; Gannon Johnson; Maziar Rostamian; Gabriel P. Potirniche; Abderrafi M. Ougouag; Massimo Bertino; Louis Franzel; Akira Tokuhiro

2013-10-01

This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

Graphite beds for coolant filtration at high temperature

International Nuclear Information System (INIS)

Heathcock, R.E.; Lacy, C.S.

1978-01-01

High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

Methodology of the On-Iine FoIIow Simulation of Pebble-bed High-temperature Reactors

International Nuclear Information System (INIS)

Xia Bing; Li Fu; Wei Chunlin; Zheng Yanhua; Chen Fubing; Zhang Jian; Guo Jiong

2014-01-01

The on-line fuel management is an essential feature of the pebble-bed high-temperature reactors (PB-HTRs), which is strongly coupled with the normal operation of the reactor. For the purpose of on-line analysis of the continuous shuffling scheme of numerous fuel pebbles, the follow simulation upon the real operation is necessary for the PB-HTRs. In this work, the on-line follow simulation methodology of the PB-HTRs’ operation is described, featured by the parallel treatments of both neutronics analysis and fuel cycling simulation. During the simulation, the operation history of the reactor is divided into a series of burn-up cycles according to the behavior of operation data, in which the steady-state neutron transport equations are solved and the diffusion theory is utilized to determine the physical features of the reactor core. The burn-up equations of heavy metals, fission products and neutron poisons including B-10, decoupled from the pebble flow term, are solved to analyze the burn-up process within a single burn-up cycle. The effect of pebble flow is simulated separately through a discrete fuel shuffling pattern confined by curved pebble flow channels, and the effect of multiple pass of the fuel is represented by logical batches within each spatial region of the core. The on-line thermal-hydraulics feedback is implemented for each bur-up cycle by using the real thermal-hydraulics data of the core operation. The treatment of control rods and absorber balls is carried out by utilizing a coupled neutron transport-diffusion calculation along with discontinuity factors. The physical models mentioned above are established mainly by using a revised version of the V.S.O.P program system. The real operation data of HTR-10 is utilized to verify the methodology presented in this work, which gives good agreement between simulation results and operation data. (author)

Effect of Different Operating Temperatures on the Biological Hydrogen Methanation in Trickle Bed Reactors

Directory of Open Access Journals (Sweden)

Andreas Lemmer

2018-05-01

Full Text Available To improve the reactor efficiency, this study investigated the influence of temperature on the biological hydrogen methanation (BHM in trickle-bed reactors (TBR. Rising temperatures increase the metabolic activity of methanogenic microorganisms, thus leading to higher reactor specific methane formation rates (MFR. In order to quantify the potential for improved performance, experiments with four different operating temperatures ranging from 40 to 55 °C were carried out. Methane content increased from 88.29 ± 2.12 vol % at 40 °C to 94.99 ± 0.81 vol % at 55 °C with a stable biological process. Furthermore, a reactor specific methane formation rate (MFR of up to 8.85 ± 0.45 m3 m−3 d−1 was achieved. It could be shown that the microorganisms were able to adapt to higher temperatures within hours. The tests showed that TBR performance with regard to BHM can be significantly increased by increasing the operating temperature.

Localization of the Hot Spot in the Gap of Pebble Bed of Very High Temperature Gas Cooled Reactor(VHTGR)

International Nuclear Information System (INIS)

Lee, Sa Ya; Hong, Sung Je; Lee, Jae Young

2010-01-01

Pebble Bed Reactor(PBR) has been investigated intensively due to its benefits in management, but its complicated flow geometry requests reliable analytical methods. Hassan and Lee et al. have been made three dimensional computational methods. Hassan also measured local velocity fields with Particle Tracking Velocimetry(PTV), in small sized packed bed using liquid coolant, and Lee et al. measured flow field in the 2-dimensional wind tunnel with a hot wire system. In the present study, we develop the scaled up wind tunnel of pebble bed to use air as coolant in the same Reynolds number condition, as 21614, of the PBMR-250MWth. In order to measure the local surface temperature, the heating system and temperature measurement system were installed and heat transfer analogy was performed. The local surface temperature data shows that the predicted hot spots by Lee et al. at the top and bottom of the pebble by the velocity field measurement are reasonable, but the heat conduction is prior than contact effect at contact points

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

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

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

Stability analysis of the high temperature thermal pebble bed nuclear reactor concept

International Nuclear Information System (INIS)

Vondy, D.R.

1981-02-01

A study was made of the stability of the high temperature gas-cooled pebble bed core against xenon-driven oscillation. This generic study indicated that a core as large as 3000 MW(t) could be stable. Several aspects present a challenge to analysis including the void space above the pebble bed, the effects of possible control rod configurations, and the temperature feedback contribution. Special methods of analysis were developed in this effort. Of considerable utility was the scheme of including an azimuthal buckling loss term in the neturon balance equations admitting direct solution of the first azimuthal harmonic for a core having azimuthal symmetry. This technique allows the linear stability analysis to be done solving two-dimensional (RZ) problems instead of three-dimensional problems. A scheme for removing the fundamental source contribution was also implemented to allow direct iteration toward the dominant harmonic solution, treating up to three dimensions with diffusion theory

Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Marmier, Alain

2012-01-01

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

Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

International Nuclear Information System (INIS)

Setiadipura, T.; Zuhair; Irwanto, D.

2015-01-01

The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

T. Setiadipura

2015-04-01

Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

Pellet bed reactor for multi-modal space power

International Nuclear Information System (INIS)

Buden, D.; Williams, K.; Mast, P.; Mims, J.

1987-01-01

A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

The high-temperature reactor

International Nuclear Information System (INIS)

Kirchner, U.

1991-01-01

The book deals with the development of the German high-temperature reactor (pebble-bed), the design of a prototype plant and its (at least provisional) shut-down in 1989. While there is a lot of material on the HTR's competitor, the fast breeder, literature is very incomplete on HTRs. The author describes HTR's history as a development which was characterised by structural divergencies but not effectively steered and monitored. There was no project-oriented 'community' such as there was for the fast breeder. Also, the new technology was difficult to control there were situations where no one quite knew what was going on. The technical conditions however were not taken as facts but as a basis for interpretation, wishes and reservations. The HTR gives an opportunity to consider the conditions under which large technical projects can be carried out today. (orig.) [de

Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)

International Nuclear Information System (INIS)

De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L.

2006-01-01

A renewed interest has been raised for liquid salt cooled nuclear reactors. The excellent heat transfer properties of liquid salt coolants provide several benefits, like lower fuel temperatures, higher coolant outlet temperatures, increased core power density and better decay heat removal. In order to benefit from the online refueling capability of a pebble bed reactor, the Liquid Salt Pebble Bed Reactor (LSPBR) is proposed. This is a high temperature pebble-bed reactor with a fuel design similar to existing HTRs, but using a liquid salt as a coolant. In this paper, the selection criteria for the liquid salt coolant are described. Based on its neutronic properties, LiF-BeF 2 (FLIBE) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic-thermal hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperatures. Finally, calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined. (authors)

Real time thermal hydraulic model for high temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

2013-01-01

A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

Programmes and projects for high-temperature reactor development

International Nuclear Information System (INIS)

Bogusch, Edgar; Hittner, Dominique

2009-01-01

An increasing attention has to be recognised worldwide on the development of High-Temperature Reactors (HTR) which has started in Germany and other countries in the 1970ies. While pebble bed reactors with spherical fuel elements have been developed and constructed in Germany, countries such as France, the US and Russia investigated HTR concepts with prismatic block-type fuel elements. The concept of a modular HTR formerly developed by Areva NP was an essential basis for the HTR-10 in China. A pebble bed HTR for electricity production is developed in South Africa. The construction is planned after the completion of the licensing procedure. Also the US is planning an HTR under the NGNP (Next Generation Nuclear Plant) Project. Due to the high temperature level of the helium coolant, the HTR can be used not only for electricity production but also for supply of process heat. Including its inherent safety features the HTR is an attractive candidate for heat supply to various types of plants e.g. for hydrogen production or coal liquefactions. The conceptual design of an HTR with prismatic fuel elements for the cogeneration of electricity and process heat has been developed by Areva NP. On the European scale the HTR development is promoted by the RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation) project. RAPHAEL is an Integrated Project of the Euratom 6th Framework Programme for the development of technologies towards a Very High-Temperature Reactor (VHTR) for the production of electricity and heat. It is financed jointly by the European Commission and the partners of the HTR Technology Network (HTR-TN) and coordinated by Areva NP. The RAPHAEL project not only promotes HTR development but also the cooperation with other European projects such as the material programme EXTREMAT. Furthermore HTR technology is investigated in the frame of Generation IV International Forum (GIF). The development of a VHTR with helium temperatures above 900 C for the

Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

International Nuclear Information System (INIS)

In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

2006-01-01

High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

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

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

International Nuclear Information System (INIS)

Peterson, Per; Greenspan, Ehud

2015-01-01

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/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

Random detailed model for probabilistic neutronic calculation in pebble bed Very High Temperature Reactors

International Nuclear Information System (INIS)

Perez Curbelo, J.; Rosales, J.; Garcia, L.; Garcia, C.; Brayner, C.

2013-01-01

The pebble bed nuclear reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed type HTRs, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. Pebble bed reactors (PBR) offer the opportunity to meet the sustainability requirements, such as nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. In order to simulate PBRs correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCPN code. The main reason is that there is a limited number of cells and surfaces to be defined. In this study, a computational tool which allows getting a new geometrical model of fuel pebbles for neutronic calculations with MCNPX code, was developed. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. Four proposed fuel pebble models were compared regarding their effective multiplication factor and energy liberation profiles. Such models are: Homogeneous Pebble, Five Zone Homogeneous Pebble, Detailed Geometry, and Randomly Detailed Geometry. (Author)

High-temperature reactor developments in the Netherlands

International Nuclear Information System (INIS)

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

1996-01-01

The high-temperature reactor development in the Netherlands 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 HTR for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will have 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. As a part of the HTR program at ECN, chemical aspects of HTR fuel and coated particles are studied. Experimental work on the oxidation resistance of coating materials and fission product attack on coating materials as well as thermochemical calculations of the fuel particles are done at ECN. The concept-HTR of ECN is fuelled with UO 2 , but the use of thorium is considered. The composition of the fuel determines the oxygen potential, which plays a key role in chemical safety of the fuel. Thermochemical calculations of the chemical form of cesium inside the HTR fuel particles were performed for a wide oxygen potential range. The chemical form of cesium determines the cesium pressure inside the fuel particle, which in turn determines the release behavior of Cs from defective particles. At normal operating temperatures and low oxygen potentials, the chemical form of cesium is C 60 Cs. It is known that cesium carbon compounds decompose above 650degC in vacuum. The stability of these compounds in the fuel particles at high temperatures(1000-1600degC) is questioned. Decomposition of these compounds may result in high cesium pressures even at normal operating conditions. Experimental work on the thermodynamic properties of cesium compounds at high temperatures is currently performed. (J.P.N.)

Evolution of Particle Bed Reactor Fuel

Science.gov (United States)

Jensen, Russell R.; Evans, Robert S.; Husser, Dewayne L.; Kerr, John M.

1994-07-01

To realize the potential performance advantages inherent in a particle bed reactor (PBR) for nuclear thermal propulsion (NTP) applications, high performance particle fuel is required. This fuel must operate safely and without failure at high temperature in high pressure, flowing hydrogen propellant. The mixed mean outlet temperature of the propellant is an important characteristic of PBR performance. This temperature is also a critical parameter for fuel particle design because it dictates the required maximum fuel operating temperature. In this paper, the evolution in PBR fuel form to achieve higher operating temperatures is discussed and the potential thermal performance of the different fuel types is evaluated. It is shown that the optimum fuel type for operation under the demanding conditions in a PBR is a coated, solid carbide particle.

Metaphysics methods development for high temperature gas cooled reactor analysis

International Nuclear Information System (INIS)

Seker, V.; Downar, T. J.

2007-01-01

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

Automated Design and Optimization of Pebble-bed Reactor Cores

International Nuclear Information System (INIS)

Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

2010-01-01

We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

The ESKOM pebble bed modular reactor

International Nuclear Information System (INIS)

Gittus, J.H.

1999-01-01

An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents

International Nuclear Information System (INIS)

2004-01-01

Description: The DRAGON Reactor Experiment (DRE): The first demonstration High temperature gas reactor (HTGR) was built in the 1960's. Thirteen OECD countries began a project in 1959 to build an experimental reactor known as Dragon at Winfrith in the UK. The reactor - which operated successfully between 1966 and 1975 - had a thermal output of 20 MW and achieved a gas outlet temperature of 750 deg. C. The High Temperature Reactor concept, if it justified its expectations, was seen as having its place as an advanced thermal reactor between the current thermal reactor types such as the PWR, BWR, and AGR and the sodium cooled fast breeder reactor. It was expected that the HTR would offer better thermal efficiency, better uranium utilisation, either with low enriched uranium fuel or high enriched uranium thorium fuel, better inherent safety and lower unit power costs. In the event all these potential advantages were demonstrated to be in principle achievable. This view is still shared today. In fact Very High Temperature Reactors is one of the concepts retained for Generation IV. Projects on constructing Modular Pebble Bed Reactors are under way. Here all available Dragon Project Reports (DPR) - approximately 1000 - are collected in electronic form. An index points to the reports (PDF format); each table in the report is accessible in EXCEL format with the aim of facilitating access to the data. These reports describe the design, experiments and modelling carried out over a period of 17 years. 2 - Related or auxiliary information: IRPHE-HTR-ARCH-01, Archive of HTR Primary Documents NEA-1728/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

International Nuclear Information System (INIS)

Boer, Brian

2007-01-01

The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

Semi-dry flue gas desulfurization using Ca(OH)2 in a fluidized bed reactor with bed materials

International Nuclear Information System (INIS)

Park, Young Oak; Roh, Hak Jae; Oh, Chang Sup; Kim, Yong Ha

2010-01-01

The main objective of present work is to reduce sulfur dioxide emission from power plant for the environment protection. The fluidized bed (FB) was used as the reactor with bed materials in a new semi-dry flue gas desulfurization (FGD) process to achieve high desulfurization efficiency (>98%). Fine powder of Ca(OH) 2 as sorbent and water were continuously fed separately to the bed reactor where bed materials (2 mm glass beads) were fluidized vigorously with flue gas (flow 720 Nm 3 / hr) using bench scale plant of stainless steel column. We have investigated different effects of water injection flow rate, Ca/ S molar ratio and weight of bed materials on SO 2 removal. The increments in the Ca/ S molar ratio and water injection flow rate have been resulted higher desulfurization efficiency with certain disadvantages such as higher sorbent cost and lower temperature of the treated flue gas, respectively. (author)

Ash behavior and de-fluidization in low temperature circulating fluidized bed biomass gasifier

DEFF Research Database (Denmark)

Narayan, Vikas

ensures that high-alkali biomass fuels can be used without risks of bed de-fluidization. This thesis aims to understand the behavior of alkali metals and ash in the LTCFB system. The thesis work involved measurements made on bed material and product gas dust samples on a 100kW LTCFB gasifier placed......Biomass is increasingly used as a fuel for power generation. Herbaceous fuels however, contain high amounts of alkali metals which get volatilized at high temperatures and forms salts with low melting points and thus condense on pipelines, reactor surfaces and may cause de-fluidization. A Low......-Temperature Circulating Fluidized Bed System (LTCFB) gasifier allows pyrolysis and gasification of biomass to occur at low temperatures thereby improving the retention of alkali and other ash species within the system and minimizing the amount of ash species in the product gas. In addition, the low reactor temperature...

Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

International Nuclear Information System (INIS)

Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

2014-01-01

The presence of water on the tube-side of the steam generators in high temperature gas-cooled reactors (HTGRs) with indirect cycle layouts presents a possibility for a penetration of neutron moderating steam into the core, which may cause a power excursion. This article presents results on the effect of water ingress into the core of the two South African Pebble Bed Modular Reactor design concepts, i.e. the PBMR-200 MW th and the PBMR-400 MW th developed by PBMR SOC Ltd. The VSOP 99/05 suite of codes was used for the simulation of this event. Partial steam vapour pressures were added in stages into the primary circuit in order to investigate the effect of water ingress on reactivity, power profiles and thermal neutron flux profiles. The effects of water ingress into the core are explained by increased neutron moderation, due to the addition of 1 H, which leads to a decrease in resonance capture by 238 U and therefore an increase in the multiplication factor. The more effective moderation of neutrons by definition reduces the fast neutron flux and increases the thermal flux in the core, i.e. leads to a softer spectrum. The more effective moderation also increases the average increase in lethargy between collisions of a neutron with successive fuel kernels, which reduces the probability for neutron capture in the radiative capture resonances of 238 U. The resulting higher resonance escape probability also increases the thermal flux in the core. The softening of the neutron spectrum leads to an increased effective microscopic fission cross section in the fissile isotopes and thus to increased neutron absorption for fission, which reduces the remaining number of neutrons that can diffuse into the reflectors. Therefore water ingress into the core leads to a reduced thermal neutron flux in the reflectors. The power density spatial distribution behaved similarly to the thermal neutron flux in the core. Analysis of possible mechanisms was conducted. The results show that

Nuclear graphite for high temperature reactors

International Nuclear Information System (INIS)

Marsden, B.J.

2001-01-01

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

The Pebble Bed Modular Reactor: An obituary

Energy Technology Data Exchange (ETDEWEB)

Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

2011-05-15

The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

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

International Nuclear Information System (INIS)

Chang Oh

2004-01-01

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

MIT pebble bed reactor project

Energy Technology Data Exchange (ETDEWEB)

Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

2007-03-15

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

MIT pebble bed reactor project

International Nuclear Information System (INIS)

Kadak, Andrew C.

2007-01-01

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

CFD-DEM simulation of a conceptual gas-cooled fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Almeida, Lucilla C.; Su, Jian

2015-01-01

Several conceptual designs of the fluidized-bed nuclear reactor have been proposed due to its many advantages over conventional nuclear reactors such as PWRs and BWRs. Amongst their characteristics, the enhanced heat transfer and mixing enables a more uniform temperature distribution, reducing the risk of hot-spot and excessive fuel temperature, in addition to resulting in a higher burnup of the fuel. Furthermore, the relationship between the bed height and reactor neutronics turns the coolant flow rate control into a power production mechanism. Moreover, the possibility of removing the fuel by gravity from the movable core in case of a loss-of-cooling accident increases its safety. High-accuracy modeling of particles and coolant flow in fluidized bed reactors is needed to evaluate reliably the thermal-hydraulic efficiency and safety margin. The two-way coupling between solid and fluid can account for high-fidelity solid-solid interaction and reasonable accuracy in fluid calculation and fluid-solid interaction. In the CFD-DEM model, the particles are modeled as a discrete phase, following the DEM approach, whereas the fluid flow is treated as a continuous phase, described by the averaged Navier-Stokes equations on a computational cell scale. In this work, the coupling methodology between Fluent and Rocky is described. The numerical approach was applied to the simulation of a bubbling fluidized bed and the results were compared to experimental data and showed good agreement. (author)

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

International Nuclear Information System (INIS)

Yoder, Graydon L.; Aaron, Adam; Cunningham, Burns; Fugate, David; Holcomb, David; Kisner, Roger; Peretz, Fred; Robb, Kevin; Wilgen, John; Wilson, Dane

2014-01-01

Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

Experimental study of flow field characteristics on bed configurations in the pebble bed reactor

International Nuclear Information System (INIS)

Jia, Xinlong; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jia, Haijun; Jiang, Shengyao

2017-01-01

Highlights: • PTV study of flow fields of pebble bed reactor with different configurations are carried out. • Some criteria are proposed to quantify vertical velocity field and flow uniformity. • The effect of different pebble bed configurations is also compared by the proposed criteria. • The displacement thickness is used analogically to analyze flow field characteristics. • The effect of mass flow variation in the stagnated region of the funnel flow is measured. - Abstract: The flow field characteristics are of fundamental importance in the design work of the pebble bed high temperature gas cooled reactor (HTGR). The different effects of bed configurations on the flow characteristics of pebble bed are studied through the PTV (Particle Tracking Velocimetry) experiment. Some criteria, e.g. flow uniformity (σ) and mass flow level (α), are proposed to estimate vertical velocity field and compare the bed configurations. The distribution of the Δθ (angle difference between the individual particle velocity and the velocity vector sum of all particles) is also used to estimate the resultant motion consistency level. Moreover, for each bed configuration, the thickness of displacement is analyzed to measure the effect of the funnel flow zone based on the boundary layer theory. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity and other characteristics; and the sequence of levels of each estimation criterion is obtained for all bed configurations. In addition, a good design of the pebble bed configuration is suggested and these estimation criteria can be also applied and adopted in testing other geometry designs of pebble bed.

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

Science.gov (United States)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

Very-high-temperature reactors for future use

International Nuclear Information System (INIS)

Kasten, P.R.

1988-08-01

Very-high-temperature reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at reactor coolant temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

International Nuclear Information System (INIS)

Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

2005-01-01

High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

Comparison of Several Thermal Conductivity Constants for Thermal Hydraulic Calculation of Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

2017-07-01

There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.

Conceptual design of a fluidized bed nuclear reactor : Statics, dynamics and safety-related aspects

NARCIS (Netherlands)

Agung, A.

2007-01-01

In this thesis a conceptual design of an innovative high temperature reactor based on the fluidization principle (FLUBER) is proposed. The reactor should satisfy the following requirements: (a) modular and low power, (b)) large shutdown margin, (c) able to produce power when the bed of particles

Pebble Bed Reactor: core physics and fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Worley, B.A.

1979-10-01

The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

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

International Nuclear Information System (INIS)

Barsell, A.W.

1980-05-01

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

Advanced Core Design And Fuel Management For Pebble-Bed Reactors

Energy Technology Data Exchange (ETDEWEB)

Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

2004-10-01

A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

Potentialities of high temperature reactors (HTR)

International Nuclear Information System (INIS)

Hittner, D.

2001-01-01

This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

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

Directory of Open Access Journals (Sweden)

Yan Wang

2017-01-01

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

Effect of bed configuration on pebble flow uniformity and stagnation in the pebble bed reactor

International Nuclear Information System (INIS)

Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jiang, Shengyao

2014-01-01

Highlights: • Pebble flow uniformity and stagnation characteristics are very important for HTR-PM. • Arc- and brachistochrone-shaped configuration effects are studied by DEM simulation. • Best bed configurations with uniform flow and no stagnated pebbles are suggested. • Detailed quantified characteristics of bed configuration effects are shown for explanation. - Abstract: Pebble flow uniformity and stagnation characteristics are very important for the design of pebble bed high temperature gas-cooled reactor. Pebble flows inside some specifically designed contraction configurations of pebble bed are studied by discrete element method. The results show the characteristics of stagnation rates, recycling rates, radial distribution of pebble velocity and residence time. It is demonstrated clearly that the bed with a brachistochrone-shaped configuration achieves optimum levels of flow uniformity and recycling rate concentration, and almost no pebbles are stagnated in the bed. Moreover, the optimum choice among the arc-shaped bed configurations is demonstrated too. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity. In addition, a good design of the pebble bed configuration is suggested

Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

International Nuclear Information System (INIS)

Hodzic, A.; Kugeler, K.

1997-01-01

This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

Optimized core design and fuel management of a pebble-bed type nuclear reactor

NARCIS (Netherlands)

Boer, B.

2009-01-01

The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

High temperature reactor development in the Netherlands

International Nuclear Information System (INIS)

Heek, A.I. van

1996-01-01

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

Study of reactivity of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Rammsy, J.E.M.

1985-01-01

The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

International Nuclear Information System (INIS)

1978-09-01

Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

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.)

Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

International Nuclear Information System (INIS)

Park, HyeongKae; Knoll, Dana; Gaston, Derek; Martineau, Richard

2010-01-01

We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

Energy Technology Data Exchange (ETDEWEB)

Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

2009-10-01

High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

High temperature energy storage performances of methane reforming with carbon dioxide in a tubular packed reactor

International Nuclear Information System (INIS)

Lu, Jianfeng; Chen, Yuan; Ding, Jing; Wang, Weilong

2016-01-01

Highlights: • Energy storage of methane reforming in a tubular packed reactor is investigated. • Thermochemical storage efficiency approaches maximum at optimal temperature. • Sensible heat and heat loss play important roles in the energy storage system. • The reaction and energy storage models of methane reforming reactor are established. • The simulated methane conversion and energy storage efficiency fit with experiments. - Abstract: High temperature heat transfer and energy storage performances of methane reforming with carbon dioxide in tubular packed reactor are investigated under different operating conditions. Experimental results show that the methane reforming in tubular packed reactor can efficiently store high temperature thermal energy, and the sensible heat and heat loss besides thermochemical energy storage play important role in the total energy storage process. When the operating temperature is increased, the thermochemical storage efficiency first increases for methane conversion rising and then decreases for heat loss rising. As the operating temperate is 800 °C, the methane conversion is 79.6%, and the thermochemical storage efficiency and total energy efficiency can be higher than 47% and 70%. According to the experimental system, the flow and reaction model of methane reforming is established using the laminar finite-rate model and Arrhenius expression, and the simulated methane conversion and energy storage efficiency fit with experimental data. Along the flow direction, the fluid temperature in the catalyst bed first decreases because of the endothermic reaction and then increases for the heat transfer from reactor wall. As a conclusion, the maximum thermochemical storage efficiency will be obtained under optimal operating temperature and optimal flow rate, and the total energy efficiency can be increased by the increase of bed conductivity and decrease of heat loss coefficient.

A new fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Sefidvash, F.

1986-01-01

A new nuclear reactor design based on the fluidized bed concept is proposed. A current design utilizes spherical fuel of slightly enriched Zircaloy-clad uranium dioxide fluidized by light water under pressure. The reactor is modular in system; therefore, any size reactor can be constructed from the basic standard modul. The reactor physics calculations show that reactivity increases with porosity to a maximum value and thereafter decreases. This produces inherent safety and eliminates the need for control rods and burnable poisons. The heat transfer calculations show that the maximum power extracted from the reactor core is not limited to the material temperature limits but to the maximum mass flow of coolant, which corresponds to the desired operating porosity. Design simplicity and inherent safety make it an attractive small reactor design. (Author) [pt

The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

International Nuclear Information System (INIS)

Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

1974-08-01

The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

The influence of thorium on the temperature reactivity coefficient in a 400 MWth pebble bed high temperature plutonium incinerating reactor

International Nuclear Information System (INIS)

Richards, Guy A.; Serfontein, Dawid E.

2014-01-01

This article investigates advanced fuel cycles containing thorium and reactor grade plutonium (Pu(PWR)) in a 400 MW th Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant. Results presented were determined from coupled neutronics and thermo-hydraulic simulations of the VSOP 99/05 diffusion codes. In a previous study impressive burn-ups (601 MWd/kg heavy metal (HM)) and thus plutonium destruction rates (69.2 %) were obtained with pure plutonium fuel with mass loadings of 3 g Pu(PWR)/fuel sphere or less. However the safety performance was poor in that the limit on the maximum fuel temperature during equilibrium operation was exceeded and positive Uniform Temperature Reactivity Coefficients (UTCs) were obtained. In the present study fuel cycles containing mixtures of thorium and plutonium achieved negative maximum UTCs. Plutonium only fuel cycles also achieved negative maximum UTCs, provided that much higher mass loadings are used. It is proposed that the lower thermal neutron flux was responsible for this effect. The plutonium only fuel cycle with 12 g Pu(PWR)/fuel sphere also achieved the adopted safety limits for the PBMR DPP-400 in that the maximum fuel temperature and the maximum power density did not exceed 1130°C or 4.5 kW/sphere respectively. This design would thus be licensable and could potentially be economically feasible. However the burn-up was much lower at 181 MWd/kgHM and thus the plutonium destruction fraction was also much lower at 24.5%, which may be sub-optimal with respect to proliferation and waste disposal objectives and therefore further optimisation studies are proposed. (author)

Critical Issues for Particle-Bed Reactor Fuels

Science.gov (United States)

Evans, Robert S.; Husser, Dewayne L.; Jensen, Russell R.; Kerr, John M.

1994-07-01

Particle-Bed Reactors (PBRs) potentially offer performance advantages for nuclear thermal propulsion, including very high power densities, thrust-to-weight ratios, and specific impulses. A key factor in achieving all of these is the development of a very-high-temperature fuel. The critical issues for all such PBR fuels are uranium loading, thermomechanical and thermochemical stability, compatibility with contacting materials, fission product retention, manufacturability, and operational tolerance for particle failures. Each issue is discussed with respect to its importance to PBR operation, its status among current fuels, and additional development needs. Mixed-carbide-based fuels are recommended for further development to support high-performance PBRs.

Particle Bed Reactor scaling relationships

International Nuclear Information System (INIS)

Slovik, G.; Araj, K.; Horn, F.L.; Ludewig, H.; Benenati, R.

1987-01-01

Scaling relationships for Particle Bed Reactors (PBRs) are discussed. The particular applications are short duration systems, i.e., for propulsion or burst power. Particle Bed Reactors can use a wide selection of different moderators and reflectors and be designed for such a wide range of power and bed power densities. Additional design considerations include the effect of varying the number of fuel elements, outlet Mach number in hot gas channel, etc. All of these variables and options result in a wide range of reactor weights and performance. Extremely light weight reactors (approximately 1 kg/MW) are possible with the appropriate choice of moderator/reflector and power density. Such systems are very attractive for propulsion systems where parasitic weight has to be minimized

High performance biological methanation in a thermophilic anaerobic trickle bed reactor.

Science.gov (United States)

Strübing, Dietmar; Huber, Bettina; Lebuhn, Michael; Drewes, Jörg E; Koch, Konrad

2017-12-01

In order to enhance energy efficiency of biological methanation of CO 2 and H 2 , this study investigated the performance of a thermophilic (55°C) anaerobic trickle bed reactor (ATBR) (58.1L) at ambient pressure. With a methane production rate of up to 15.4m 3 CH4 /(m 3 trickle bed ·d) at methane concentrations above 98%, the ATBR can easily compete with the performance of other mixed culture methanation reactors. Control of pH and nutrient supply turned out to be crucial for stable operation and was affected significantly by dilution due to metabolic water production, especially during demand-orientated operation. Considering practical applications, inoculation with digested sludge, containing a diverse biocenosis, showed high adaptive capacity due to intrinsic biological diversity. However, no macroscopic biofilm formation was observed at thermophilic conditions even after 313days of operation. The applied approach illustrates the high potential of thermophilic ATBRs as a very efficient energy conversion and storage technology. Copyright © 2017 Elsevier Ltd. All rights reserved.

Oxidation of ethene in a wall-cooled packed-bed reactor

NARCIS (Netherlands)

Schouten, E.P.S.; Borman, P.C.; Westerterp, K.R.

1994-01-01

The selective oxidation of ethene over a silver on α-alumina catalyst was studied in a pilot plant with a wall-cooled tubular packed bed reactor. Gas and solid temperatures in the catalyst bed were measured at different axial and radial positions as well as concentrations at different axial

Passive cooling of a fixed bed nuclear reactor

International Nuclear Information System (INIS)

Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

2005-01-01

Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

CFD study on the supercritical carbon dioxide cooled pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Yu, Dali, E-mail: ydlmitd@outlook.com; Peng, Minjun; Wang, Zhongyi

2015-01-15

Highlights: • An innovation concept of supercritical carbon dioxide cooled pebble bed reactor is proposed. • Body-centered cuboid (BCCa) arrangement is adopted for the pebbles. • S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor. - Abstract: The thermal hydraulic study of using supercritical carbon dioxide (S-CO{sub 2}), a superior fluid state brayton cycle medium, in pebble bed type nuclear reactor is assessed through computational fluid dynamics (CFD) methodology. Preliminary concept design of this S-CO{sub 2} cooled pebble bed reactor (PBR) is implemented by the well-known KTA heat transfer correlation and Ergun pressure drop equation. Eddy viscosity transport turbulence model is adopted and verified by KTA calculated results. Distributions of the temperature, velocity, pressure and Nusselt (Nu) number of the coolant near the surface of the middle spherical fuel element are obtained and analyzed. The conclusion of the assessment is that S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor due primarily to its good heat transfer characteristic and large mass density, which could lead to achieve lower pressure drop and higher power density.

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

International Nuclear Information System (INIS)

Hart, R.S.; Kendall, J.M.; Marsden, B.J.

2003-01-01

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 MW th 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

Experimental investigation of thermal de-stratification in rock bed TES systems for high temperature applications

International Nuclear Information System (INIS)

Okello, Denis; Nydal, Ole J.; Banda, Eldad J.K.

2014-01-01

Highlights: • High thermal stratifications exists rock bed TES when charge with high temperature heat. • Faster thermal degradation occurs in highly stratified bed irrespective of the bed length. • Average rate of heat loss as a function of storage time increases with increasing average bed temperature. - Abstract: Solar energy fluctuates so much that it cannot promote continuous use. Integration of Thermal Energy Storage (TES) with solar energy collection devices has the potential of making solar energy available on demand. Thermal energy can be stored in a bed of rocks at temperatures suitable for applications like cooking, boiling space heating, etc. During charging, temperature stratification is observed in the bed. In a stratified system, if the heat is used immediately, then it is possible to extract heat at reasonably high temperature from the top. For cases where the system is to be used after sometime (later at night or the following morning), the high temperature heat at the top is observed to degrade as the system tries to establish thermal equilibrium irrespective of the bed height. The average rate of heat loss from the TES unit to the ambient is found to increase with increasing average bed temperatures

Esterification of oleic acid in a three-phase, fixed-bed reactor packed with a cation exchange resin catalyst.

Science.gov (United States)

Son, Sung Mo; Kimura, Hiroko; Kusakabe, Katsuki

2011-01-01

Esterification of oleic acid was performed in a three-phase fixed-bed reactor with a cation exchange resin catalyst (Amberlyst-15) at high temperature, which was varied from 80 to 120 °C. The fatty acid methyl ester (FAME) yields in the fixed-bed reactor were increased with increases in the reaction temperature, methanol flow rate and bed height. Moreover, the FAME yields were higher than those obtained using a batch reactor due to an equilibrium shift toward the product that resulted from continuous evaporation of the produced water. In addition, there was no catalyst deactivation during the esterification of oleic acid. However, addition of sunflower oil to the oleic acid reduced the FAME yield obtained from simultaneous esterification and transesterification. The FAME yield was 97.5% at a reaction temperature of 100 °C in the fixed-bed with a height of 5 cm when the methanol and oleic acid feed rates were 8.6 and 9.0 mL/h, respectively. Copyright © 2010 Elsevier Ltd. All rights reserved.

Methanol steam-reforming in a catalytic fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Duesterwald, H G; Hoehlein, B; Kraut, H; Meusinger, J; Peters, R [Research Centre Juelich (KFA) (Germany). Inst. of Energy Process Engineering; Stimming, U [Technische Univ. Muenchen, Garching (Germany). Inst. fuer Festkoerperphysik und Techn. Phys.

1997-12-01

Designing an appropriate methanol steam reformer requires detailed knowledge about the processes within such a reactor. Thus, the axial temperature and concentration gradients and catalyst ageing were investigated. It was found that for a fresh catalyst load, the catalyst located in the reactor entrance was most active during the experiment. The activity of this part of the catalyst bed decreased after some time of operation due to ageing. With further operation, the most active zone moved through the catalyst bed. From the results concerning hydrogen production and catalyst degradation, the necessary amount of catalyst for a mobile PEMFC-system can be estimated. (orig.)

Celebrating 40 years anaerobic sludge bed reactors for industrial wastewater treatment

NARCIS (Netherlands)

Van Lier, J.B.; Van der Zee, F.P.; Frijters, C.T.M.J.; Ersahin, M.E.

2015-01-01

In the last 40 years, anaerobic sludge bed reactor technology evolved from localized lab-scale trials to worldwide successful implementations at a variety of industries. High-rate sludge bed reactors are characterized by a very small foot print and high applicable volumetric loading rates. Best

Lagrangian Approach to Study Catalytic Fluidized Bed Reactors

Science.gov (United States)

Madi, Hossein; Hossein Madi Team; Marcelo Kaufman Rechulski Collaboration; Christian Ludwig Collaboration; Tilman Schildhauer Collaboration

2013-03-01

Lagrangian approach of fluidized bed reactors is a method, which simulates the movement of catalyst particles (caused by the fluidization) by changing the gas composition around them. Application of such an investigation is in the analysis of the state of catalysts and surface reactions under quasi-operando conditions. The hydrodynamics of catalyst particles within a fluidized bed reactor was studied to improve a Lagrangian approach. A fluidized bed methanation employed in the production of Synthetic Natural Gas from wood was chosen as the case study. The Lagrangian perspective was modified and improved to include different particle circulation patterns, which were investigated through this study. Experiments were designed to evaluate the concepts of the model. The results indicate that the setup is able to perform the designed experiments and a good agreement between the simulation and the experimental results were observed. It has been shown that fluidized bed reactors, as opposed to fixed beds, can be used to avoid the deactivation of the methanation catalyst due to carbon deposits. Carbon deposition on the catalysts tested with the Lagrangian approach was investigated by temperature programmed oxidation (TPO) analysis of ex-situ catalyst samples. This investigation was done to identify the effects of particles velocity and their circulation patterns on the amount and type of deposited carbon on the catalyst surface. Ecole Polytechnique Federale de Lausanne(EPFL), Paul Scherrer Institute (PSI)

Process heat cogeneration using a high temperature reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

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.

A high temperature reactor for ship propulsion

International Nuclear Information System (INIS)

Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

2002-01-01

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

Modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shepherd, L.R.

1988-01-01

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

Benchmark Analysis Of The High Temperature Gas Cooled Reactors Using Monte Carlo Technique

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Huda, M.Q.

2008-01-01

Information about several past and present experimental and prototypical facilities based on High Temperature Gas-Cooled Reactor (HTGR) concepts have been examined to assess the potential of these facilities for use in this benchmarking effort. Both reactors and critical facilities applicable to pebble-bed type cores have been considered. Two facilities - HTR-PROTEUS of Switzerland and HTR-10 of China and one conceptual design from Germany - HTR-PAP20 - appear to have the greatest potential for use in benchmarking the codes. This study presents the benchmark analysis of these reactors technologies by using MCNP4C2 and MVP/GMVP Codes to support the evaluation and future development of HTGRs. The ultimate objective of this work is to identify and develop new capabilities needed to support Generation IV initiative. (author)

Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

International Nuclear Information System (INIS)

1978-09-01

Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

Pellet bed reactor for nuclear thermal propelled vehicles

International Nuclear Information System (INIS)

El-Genk, M.; Morley, N.J.; Haloulakos, V.E.

1991-01-01

The Pellet Bed Reactor (PeBR) concept is capable of operating at a high power density of up to 3.0 kWt/cu cm and an exit hydrogen gas temperature of 3000 K. The nominal reactor thermal power is 1500 MW and the reactor core is 0.80 m in diameter and 1.3 m high. The nominal PeBR engine generates a thrust of approximately 315 kN at a specific impulse of 1000 s for a mission duration to Mars of 250 days requiring a total firing time of 170 minutes. Because of its low diameter-to-height ratio, PeBR has enough surface area for passive removal of the decay heat from the reactor core. The reactor is equipped with two independent shutdown mechanisms; 8-B4C safety rods and 26 BeO/B4C control drums; each system is capable of operating and scraming the reactor safely. Due to the absence of core internal support structures, the PeBR can be fueled and refueled in orbit using the vacuum of space. These unique features of the PeBR provide for safety during launch, simplicity of handling, deployment, and end-of-life disposal, and vehicle extended lifetime. 11 refs

Operation of a steam hydro-gasifier in a fluidized bed reactor

OpenAIRE

Park, Chan Seung; Norbeck, Joseph N.

2008-01-01

Carbonaceous material, which can comprise municipal waste, biomass, wood, coal, or a natural or synthetic polymer, is converted to a stream of methane and carbon monoxide rich gas by heating the carbonaceous material in a fluidized bed reactor using hydrogen, as fluidizing medium, and using steam, under reducing conditions at a temperature and pressure sufficient to generate a stream of methane and carbon monoxide rich gas but at a temperature low enough and/or at a pressure high enough to en...

Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

International Nuclear Information System (INIS)

Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

2009-01-01

Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

Rotating-bed reactor as a power source for EM gun applications

Energy Technology Data Exchange (ETDEWEB)

Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

1980-01-01

Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

Rotating-bed reactor as a power source for EM gun applications

International Nuclear Information System (INIS)

Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

1980-01-01

Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m 3 ), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300 0 K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications

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

International Nuclear Information System (INIS)

Cole, Quentin

2013-01-01

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

High temperature reactor safety and environment

International Nuclear Information System (INIS)

Brisbois, J.; Charles, J.

1975-01-01

High-temperature reactors are endowed with favorable safety and environmental factors resulting from inherent design, main-component safety margins, and conventional safety systems. The combination of such characteristics, along with high yields, prove in addition, that such reactors are plagued with few problems, can be installed near users, and broaden the recourse to specific power, therefore fitting well within a natural environment [fr

Very-high-temperature reactors for future use

International Nuclear Information System (INIS)

Kasten, P.R.

1988-01-01

Very-High-Temperature Reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at helium temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

High temperature gas cooled nuclear reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

Fluid flow and heat transfer investigation of pebble bed reactors using mesh-adaptive LES

International Nuclear Information System (INIS)

Pavlidis, Dimitrios; Lathouwers, Danny

2013-01-01

The very high temperature reactor is one of the designs currently being considered for nuclear power generation. One its variants is the pebble bed reactor in which the coolant passes through complex geometries (pores) at high Reynolds numbers. A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in such reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. coolant flow and heat transfer patterns are investigated

A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

Energy Technology Data Exchange (ETDEWEB)

Moormann, R.

2008-06-15

The AVR pebble bed reactor (46 MW{sub th}) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW{sub th}). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were

A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

International Nuclear Information System (INIS)

Moormann, R.

2008-06-01

The AVR pebble bed reactor (46 MW th ) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW th ). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were observed

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

Theory and measurements of electrophoretic effects in monolith, fixed-bed, and fluidized-bed plasma reactors

International Nuclear Information System (INIS)

Morin, T.J.

1989-01-01

Pressure gradients and secondary flow fields generated by the passage of electrical current in a d.c. gas discharge or gas laser are topics of longstanding interest in the gaseous electronics literature. These hydrodynamic effects of space charge fields and charged particle density gradients have been principally exploited in the development of gas separation and purification processes. In recent characterization studies of fixed-bed and fluidized-bed plasma reactors several anomalous flow features have been observed. These reactors involve the contacting of a high-frequency, resonantly-sustained, disperse gas discharge with granular solids in a fixed or fluidized bed. Anomalies in the measured pressure drops and fluidization velocities have motivated the development of an appropriate theoretical approach to, and some additional experimental investigations of electrophoretic effects in disperse gas discharges. In this paper, a theory which includes the effects of space charge and diffusion is used to estimate the electric field and charged particle density profiles. These profiles are then used to calculate velocity fields and gas flow rates for monolith, fixed-bed, and fluidized-bed reactors. These results are used to rationalize measurements of gas flow rates and axial pressure gradients in high-frequency disperse gas discharges with and without an additional d.c. axial electric field

Characterization of fluidization regime in circulating fluidized bed reactor with high solid particle concentration using computational fluid dynamics

Energy Technology Data Exchange (ETDEWEB)

Chalermsinsuwan, Benjapon; Thummakul, Theeranan; Piumsomboon, Pornpote [Chulalongkorn University, Bangkok (Thailand); Gidaspow, Dimitri [Armour College of Engineering, Chicago (United States)

2014-02-15

The hydrodynamics inside a high solid particle concentration circulating fluidized bed reactor was investigated using computational fluid dynamics simulation. Compared to a low solid particle reactor, all the conventional fluidization regimes were observed. In addition, two unconventional fluidization regimes, circulating-turbulent and dense suspension bypassing regimes, were found with only primary gas injection. The circulating-turbulent fluidization regime showed uniformly dense solid particle distribution in all the system directions, while the dense suspension bypassing fluidization regime exhibited the flow of solid particles at only one side system wall. Then, comprehensive fluidization regime clarification and mapping were evaluated using in-depth system parameters. In the circulating-turbulent fluidization regime, the total granular temperature was low compared to the adjacent fluidization regimes. In the dense suspension bypassing fluidization regime, the highest total granular temperature was obtained. The circulating-turbulent and dense suspension bypassing fluidization regimes are suitable for sorption and transportation applications, respectively.

DEGRADATION OF AROMATIC COMPOUNDS USING MOVING BED BIOFILM REACTORS

Directory of Open Access Journals (Sweden)

B. Ayati, H. Ganjidoust, M. Mir Fattah

2007-04-01

Full Text Available For biological treatment of water, there are many different biofilm systems in use. Examples of them are trickling filters, rotating biological contactors, fixed media submerged biofilters, granular media biofilters and fluidized bed reactors. They all have their advantages and disadvantages. Hence, the Moving Bed Biofilm Reactor process was developed in Norway in the late 1980s and early 1990s to adopt the best features of the activated sludge process as well as those of the biofilter processes, without including the worst. Two cylindrical moving bed biofilm reactors were used in this study working in upflow stream conditions. Experiments have been done in aerobic batch flow regime. Laboratory experiments were conducted at room temperature (23–28C and synthetic wastewater comprising a composition of phenol and hydroquinone in each reactor as the main organic constituents, plus balanced nutrients and alkalinity were used to feed the reactor. The ratio of influent to effluent COD was determined at different retention times. The results indicated that the removal efficiency of each selected compound is affected by the detention time. At low phenol and hydroquinone concentration (from 700 to 1000 mg/L maximum removal efficiency (over 80 % was obtained. By further increasing in COD loading rate up to 3000 mg/L, a decrease in COD removal rate was occurred. In the reactor containing pyrogallol in COD of 1500 mg/L, the removal rate decreased to 10 percent because of its toxicity for microorganisms.

Nuclear reactor application for high temperature power industrial processes

International Nuclear Information System (INIS)

Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

1977-01-01

This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

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

1985-11-01

This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

Recent improvements in the filtration of corrosion products in high temperature water and application to reactor circuits

International Nuclear Information System (INIS)

Darras, R.; Dolle, L.; Chenouard, J.; Laylavoix, F.

1977-01-01

The nature and physico-chemical behavior of corrosion products released by structural materials into high temperature water flowing in power reactor circuits have been investigated in test loops and different power plants. The results improve more particularly the knowledge of probable rate constants governing their disappearance through deposition of crud on the fuel cladding. It appears that a considerable limitation of radioactivity transportation in the primary circuit components of pressurized water reactors is in a general way only possible through extraction of the corrosion products by filtration at a rate adequate to minimize the amount of crud deposited in the core. This extraction rate has been estimated; its magnitude implicates a filtration operating on the high temperature water in the primary circuit which allows the necessary high flows. The application of magnetic and electromagnetic so as deep granular graphite bed filters has been studied. The results concerning efficiencies and limiting yields at high temperatures are given. Estimates concerning technological feasibility and corresponding investments are discussed

Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

International Nuclear Information System (INIS)

In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

2008-01-01

Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

Hydrodynamics of multi-phase packed bed micro-reactors

NARCIS (Netherlands)

Márquez Luzardo, N.M.

2010-01-01

Why to use packed bed micro-reactors for catalyst testing? Miniaturized packed bed reactors have a large surface-to-volume ratio at the reactor and particle level that favors the heat- and mass-transfer processes at all scales (intra-particle, inter-phase and inter-particle or reactor level). If the

Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

Directory of Open Access Journals (Sweden)

Hammam Oktajianto

2014-12-01

Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

Design and development of fluidized bed reactor system for production of trichlorosilane as a precursor for high purity silicon

International Nuclear Information System (INIS)

Kumar, Rajesh; Mohan, Sadhana; Bhanja, K.; Nayak, S.; Bhattacharya, S.K.

2009-01-01

Trichlorosilane is widely used as precursor material for production of high purity silicon. It is mainly produced by reaction of metallurgical grade silicon with anhydrous HCl gas in a fluidized bed reactor. To develop this process on commercial scale a pilot size fluidized bed reactor system was designed and developed and successfully operated. This paper discusses the critical issues related to these activities. (author)

Effect of friction on pebble flow pattern in pebble bed reactor

International Nuclear Information System (INIS)

Li, Yu; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jiang, Shengyao

2016-01-01

Highlights: • A 3D DEM study on particle–wall/particle friction in pebble bed reactor is carried out. • Characteristic values are defined to evaluate features of pebble flow pattern quantitatively. • Particle–wall friction is dominant to determine flow pattern in a specific pebble bed. • Friction effect of hopper part on flow field is more critical than that of cylinder part. • Three cases of 1:1 full scale practical pebble beds are simulated for demonstration. - Abstract: Friction affects pebble flow pattern in pebble-bed high temperature gas-cooled reactor (HTGR) significantly. Through a series of three dimensional DEM (discrete element method) simulations it is shown that reducing friction can be beneficial and create a uniform and consistent flow field required by nuclear engineering. Particle–wall friction poses a decisive impact on flow pattern, and particle–particle friction usually plays a secondary role; relation between particle–wall friction and flow pattern transition is also concluded. Moreover, new criteria are created to describe flow patterns quantitatively according to crucial issues in HTGR like stagnant zone, radial uniformity and flow sequence. Last but not least, it is proved that friction control of hopper part is more important than that of cylinder part in practical pebble beds, so reducing friction between pebbles and hopper surface is the engineering priority.

Comparison of packed bed and fluidized bed membrane reactors for methane reforming

NARCIS (Netherlands)

Gallucci, F.; van Sint Annaland, M.; Kuipers, J.A.M.

2009-01-01

In this work the performance of different membrane reactor concepts, both fluidized bed and packed bed membrane reactors, have been compared for the reforming of methane for the production of ultra-pure hydrogen. Using detailed theoretical models, the required membrane area to reach a given

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

Modeling of reaction kinetics in bubbling fluidized bed biomass gasification reactor

Energy Technology Data Exchange (ETDEWEB)

Thapa, R.K.; Halvorsen, B.M. [Telemark University College, Kjolnes ring 56, P.O. Box 203, 3901 Porsgrunn (Norway); Pfeifer, C. [University of Natural Resources and Life Sciences, Vienna (Austria)

2013-07-01

Bubbling fluidized beds are widely used as biomass gasification reactors as at the biomass gasification plant in Gussing, Austria. The reactor in the plant is a dual circulating bubbling fluidized bed gasification reactor. The plant produces 2MW electricity and 4.5MW heat from the gasification of biomass. Wood chips as biomass and olivine particles as hot bed materials are fluidized with high temperature steam in the reactor. As a result, biomass undergoes endothermic chemical reaction to produce a mixture of combustible gases in addition to some carbon-dioxide (CO2). The combustible gases are mainly hydrogen (H2), carbon monoxide (CO) and methane (CH4). The gas is used to produce electricity and heat via utilization in a gas engine. Alternatively, the gas is further processed for gaseous or liquid fuels, but still on the process of development level. Composition and quality of the gas determine the efficiency of the reactor. A computational model has been developed for the study of reaction kinetics in the gasification rector. The simulation is performed using commercial software Barracuda virtual reactor, VR15. Eulerian-Lagrangian approach in coupling of gas-solid flow has been implemented. Fluid phase is treated with an Eulerian formulation. Discrete phase is treated with a Lagrangian formulation. Particle-particle and particle-wall interactions and inter-phase heat and mass transfer have been taken into account. Series of simulations have been performed to study model prediction of the gas composition. The composition is compared with data from the gasifier at the CHP plant in Güssing, Austria. The model prediction of the composition of gases has good agreements with the result of the operating plant.

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

Energy Technology Data Exchange (ETDEWEB)

Brian Boer; Abderrafi M. Ougouag

2011-03-01

The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

International Nuclear Information System (INIS)

Boer, Brian; Ougouag, Abderrafi M.

2011-01-01

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant

Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2015-01-01

Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

Coal-char combustion in a fluidised bed reactor

Energy Technology Data Exchange (ETDEWEB)

Mehrotra, S.P.; Pande, M. [Indian Institute of Technolgy, Kanpur (India)

2001-12-01

Combustion of bituminous coal chars ranging from 0.8 mm to 1.8 mm has been studied in a fluidised bed reactor at temperatures ranging from 500 to 850{sup o}C. The fluidised bed consists of inert sand particles of average size of 0.5 mm and reactive coal char particles. A heat balance has been worked out to calculate the rate of combustion of char from measured incremental changes in the bed temperature during combustion. Investigations on partially burnt particles suggest that the ash layer which builds up around the burning core of char particles is non-flaking and the particles burn in a shrinking core manner. Analysis of rate data indicates that the rate of combustion is controlled by chemical reaction kinetics, though diffusion of oxygen through the bundary layer begins to influence the overall reaction kinetics at higher temperatures. The burnt out time varies linearly with particle size. Activation energy for the chemical reaction control regime is found to be around 68 kJ/mole.

Physics of the pebble-bed high temperature reactor in massive water ingress accidents

International Nuclear Information System (INIS)

Scherer, W.

1989-10-01

A point-kinetics model was developed to describe qualitatively hypothetical water ingress transients in the primary loop of High Temperature Reactors. Neutron kinetics, heat-flow balance and the chemical reaction of graphite corrosion together with their mutual influence are included. The qualitative behaviour of the transients is calculated and discussed for two fictitious examples, namely the long-term water ingress into a medium sized HTR (HTR-500) and the 'startup' of a small HTR after an intensive water flooding of the core. The model developed and the computer code KINKOR are thought to be tools for the general understanding of the water ingress phenomena and should be looked at as basis for more elaborated systems. (orig./HP) [de

Fixed-bed Reactor Dynamics and Control - A Review

DEFF Research Database (Denmark)

Jørgensen, S. B.

1986-01-01

The industrial diversity of fixed bed reactors offers a challenging and relevant set of control problems. These intricate problems arise due to the rather complex dynamics of fixed bed reactors and to the complexity of actual reactor configurations. Many of these control problems are nonlinear...... and multi-variable. During the last decade fixed bed reactor control strategies have been proposed and investigated experimentally. This paper reviews research on these complex control problems with an emphasis upon solutions which have been demon-strated to work in the laboratory and hold promise...

Study on the profitableness of electricity generation with high temperature reactors

International Nuclear Information System (INIS)

Kolb, G.

1978-08-01

The programme group 'Systemforschung und Technologische Entwicklung' (STE) of the Nuclear Research Centre Juelich in cooperation with the internal institutions 'Projekttraegerschaft Entwicklung von Hochtemperaturreaktoren' (PTH) and 'Institut fuer Reaktorentwicklung' (IRE) on the one hand, and the external partner 'Hochtemperatur-Reactorbau GmbH' (HRB) and 'Gesellschaft fuer Hochtemperaturreactor Technik' (GHT) on the other hand have set up a study on fuel cycle costs, electricity production cost and the economical use as well as uranium resource protection by introduction of high temperature reactors (HTR) with pebble bed core to generate electricity. The pressurized-water reactor (PWR) today on the market serves as comparison. The working results obtained sofar are compiled in the present report. It was particularly noted that - the HTR can economically fully compete with the PWR for electricity generation - the necessary supply of natural uranium for the HTR in open circuit is about one third lower and in the closed circuit, almost two thirds lower than in the corresponding PWR. A further reduction is possible on a long-term basis by highly converting HTW systems. (orig.) [de

Pebble-bed reactor

International Nuclear Information System (INIS)

Lohnert, G.; Mueller-Frank, U.; Heil, J.

1976-01-01

A pebble-bed nuclear reactor of large power rating comprises a container having a funnel-shaped bottom forming a pebble run-out having a centrally positioned outlet. A bed of downwardly-flowing substantially spherical nuclear fuel pebbles is positioned in the container and forms a reactive nuclear core maintained by feeding unused pebbles to the bed's top surface while used or burned-out pebbles run out and discharge through the outlet. A substantially conical body with its apex pointing upwardly and its periphery spaced from the periphery of the container spreads the bottom of the bed outwardly to provide an annular flow down the funnel-shaped bottom forming the runout, to the discharge outlet. This provides a largely constant downward velocity of the spheres throughout the diameter of the bed throughout a substantial portion of the down travel, so that all spheres reach about the same burned-out condition when they leave the core, after a single pass through the core area

Carbon dioxide hydrate formation in a fixed-bed reactor

Energy Technology Data Exchange (ETDEWEB)

Fan, S.; Lang, X. [South China Univ. of Technology, Guangzhou (China). Key Laboratory of Enhanced Heat Transfer and Energy Conservation; Wang, Y.; Liang, D. [Chinese Academy of Sciences, Guangzhou (China). Guangzhou Inst. of Energy Conversion and Guangzhou Center of Natural Gas Hydrate; Sun, X.; Jurcik, B. [Air Liquide Laboratories, Tsukuba (Japan)

2008-07-01

Gas hydrates are thermodynamically stable at high pressures and near the freezing temperature of pure water. Methane hydrates occur naturally in sediments in the deep oceans and permafrost regions and constitute an extensive hydrocarbon reservoir. Carbon dioxide (CO{sub 2}) hydrates are of interest as a medium for marine sequestration of anthropogenic carbon dioxide. Sequestering CO{sub 2} as hydrate has potential advantages over most methods proposed for marine CO{sub 2} sequestration. Because this technique requires a shallower depth of injection when compared with other ocean sequestration methods, the costs of CO{sub 2} hydrate sequestration may be lower. Many studies have successfully used different continuous reactor designs to produce CO{sub 2} hydrates in both laboratory and field settings. This paper discussed a study that involved the design and construction of a fixed-bed reactor for simulation of hydrate formation system. Water, river sands and carbon dioxide were used to simulate the seep kind of hydrate formation. Carbon dioxide gas was distributed as small bubbles to enter from the bottom of the fixed-bed reactor. The paper discussed the experimental data and presented a diagram of the gas hydrate reactor system. The morphology as well as the reaction characters of CO{sub 2} hydrate was presented in detail. The results were discussed in terms of experimental phenomena and hydrate formation rate. A mathematical model was proposed for describing the process. 17 refs., 7 figs.

Fluidized Bed Reactor as Solid State Fermenter

Directory of Open Access Journals (Sweden)

Krishnaiah, K.

2005-01-01

Full Text Available Various reactors such as tray, packed bed, rotating drum can be used for solid-state fermentation. In this paper the possibility of fluidized bed reactor as solid-state fermenter is considered. The design parameters, which affect the performances are identified and discussed. This information, in general can be used in the design and the development of an efficient fluidized bed solid-state fermenter. However, the objective here is to develop fluidized bed solid-state fermenter for palm kernel cake conversion into enriched animal and poultry feed.

Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

International Nuclear Information System (INIS)

Dahl, Pamela M.; Su, Jian

2017-01-01

Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Dahl, Pamela M.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

Steam reforming of heptane in a fluidized bed membrane reactor

Science.gov (United States)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

High Temperature Fluoride Salt Test Loop

Energy Technology Data Exchange (ETDEWEB)

Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-01

Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels

Fluidized-bed reactors processes and operating conditions

CERN Document Server

Yates, John G

2016-01-01

The fluidized-bed reactor is the centerpiece of industrial fluidization processes. This book focuses on the design and operation of fluidized beds in many different industrial processes, emphasizing the rationale for choosing fluidized beds for each particular process. The book starts with a brief history of fluidization from its inception in the 1940’s. The authors present both the fluid dynamics of gas-solid fluidized beds and the extensive experimental studies of operating systems and they set them in the context of operating processes that use fluid-bed reactors. Chemical engineering students and postdocs as well as practicing engineers will find great interest in this book.

Catalytic wet oxidation of phenol in a trickle bed reactor over a Pt/TiO2 catalyst.

Science.gov (United States)

Maugans, Clayton B; Akgerman, Aydin

2003-01-01

Catalytic wet oxidation of phenol was studied in a batch and a trickle bed reactor using 4.45% Pt/TiO2 catalyst in the temperature range 150-205 degrees C. Kinetic data were obtained from batch reactor studies and used to model the reaction kinetics for phenol disappearance and for total organic carbon disappearance. Trickle bed experiments were then performed to generate data from a heterogeneous flow reactor. Catalyst deactivation was observed in the trickle bed reactor, although the exact cause was not determined. Deactivation was observed to linearly increase with the cumulative amount of phenol that had passed over the catalyst bed. Trickle bed reactor modeling was performed using a three-phase heterogeneous model. Model parameters were determined from literature correlations, batch derived kinetic data, and trickle bed derived catalyst deactivation data. The model equations were solved using orthogonal collocations on finite elements. Trickle bed performance was successfully predicted using the batch derived kinetic model and the three-phase reactor model. Thus, using the kinetics determined from limited data in the batch mode, it is possible to predict continuous flow multiphase reactor performance.

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

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

International Nuclear Information System (INIS)

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

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.

Design of high temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

1994-09-01

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)

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

International Nuclear Information System (INIS)

Snead, Lance Lewis; Besmann, Theodore M.; Collins, Emory D.; Bell, Gary L.

2010-01-01

The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

Theoretical comparison of packed bed and fluidized bed membrane reactors for methane reforming

NARCIS (Netherlands)

Gallucci, F.; van Sint Annaland, M.; Kuipers, J.A.M.

2010-01-01

In this theoretical work the performance of different membrane reactor concepts, both fluidized bed and packed bed membrane reactors, has been compared for ultra-pure hydrogen production via methane reforming. Using detailed theoretical models, the required membrane area to reach a given conversion

Integrated design approach of the pebble BeD modular reactor using models

International Nuclear Information System (INIS)

Venter, Pieter J.; Mitchell, Mark N.

2007-01-01

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR

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

International Nuclear Information System (INIS)

Jia Qianqian; Qu Ronghong; Zhang Liangju

2014-01-01

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

Pyrolysis of softwood carbohydrates in a fluidized bed reactor.

Science.gov (United States)

Aho, Atte; Kumar, Narendra; Eränen, Kari; Holmbom, Bjarne; Hupa, Mikko; Salmi, Tapio; Murzin, Dmitry Yu

2008-09-01

In the present work pyrolysis of pure pine wood and softwood carbohydrates, namely cellulose and galactoglucomannan (the major hemicellulose in coniferous wood), was conducted in a batch mode operated fluidized bed reactor. Temperature ramping (5 degrees C/min) was applied to the heating until a reactor temperature of 460 degrees C was reached. Thereafter the temperature was kept until the release of non-condensable gases stopped. The different raw materials gave significantly different bio-oils. Levoglucosan was the dominant product in the cellulose pyrolysis oil. Acetic acid was found in the highest concentrations in both the galactoglucomannan and in the pine wood pyrolysis oils. Acetic acid is most likely formed by removal of O-acetyl groups from mannose units present in GGM structure.

High temperature brazing of reactor materials

International Nuclear Information System (INIS)

Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

1990-01-01

Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

High temperature fast reactor for hydrogen production in Brazil

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

2008-01-01

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)

User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

1982-03-01

The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

User's manual for ASTERIX-2: a two-dimensional modular-code system for the steady-state and xenon-transient analysis of a pebble-bed high-temperature reactor

International Nuclear Information System (INIS)

Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.

1982-03-01

The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution

Improved lignin pyrolysis for phenolics production in a bubbling bed reactor--Effect of bed materials.

Science.gov (United States)

Li, Dongbing; Briens, Cedric; Berruti, Franco

2015-01-01

Lignin pyrolysis was studied in a bubbling fluidized bed reactor equipped with a fractional condensation train, using nitrogen as the fluidization gas. The effect of different bed materials (silica sand, lignin char, activated lignin char, birch bark char, and foamed glass beads) on bio-oil yield and quality was investigated for a pyrolysis temperature of 550 °C. Results how that a bed of activated lignin char is preferable to the commonly used silica sand: pyrolysis of Kraft lignin with a bed of activated lignin char not only provides a pure char product, but also a higher dry bio-oil yield (with a relative increase of 43%), lower pyrolytic water production, and better bio-oil quality. The bio-oil obtained from Kraft lignin pyrolysis with a bed of activated lignin char has a lower average molecular weight, less tar, more phenolics, and less acidity than when sand is used as bed material. Copyright © 2015 Elsevier Ltd. All rights reserved.

Resonance integral calculations for high temperature reactors

International Nuclear Information System (INIS)

Blake, J.P.H.

1960-02-01

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)

Method and apparatus for a combination moving bed thermal treatment reactor and moving bed filter

Energy Technology Data Exchange (ETDEWEB)

Badger, Phillip C.; Dunn, Jr., Kenneth J.

2015-09-01

A moving bed gasification/thermal treatment reactor includes a geometry in which moving bed reactor particles serve as both a moving bed filter and a heat carrier to provide thermal energy for thermal treatment reactions, such that the moving bed filter and the heat carrier are one and the same to remove solid particulates or droplets generated by thermal treatment processes or injected into the moving bed filter from other sources.

Gas cooled thermal reactors with high temperatures (VHTR)

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.

2014-01-01

VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)

Operation of Packed-Bed Reactors Studied in Microgravity

Science.gov (United States)

Motil, Brian J.; Balakotaiah, Vemuri

2004-01-01

The operation of a packed bed reactor (PBR) involves gas and liquid flowing simultaneously through a fixed-bed of solid particles. Depending on the application, the particles can be various shapes and sizes but are generally designed to force the two fluid phases through a tortuous route of narrow channels connecting the interstitial space. The PBR is the most common type of reactor in industry because it provides for intimate contact and high rates of transport between the phases needed to sustain chemical or biological reactions. The packing may also serve as either a catalyst or as a support for growing biological material. Furthermore, this type of reactor is relatively compact and requires minimal power to operate. This makes it an excellent candidate for unit operations in support of long-duration human space activities.

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

International Nuclear Information System (INIS)

Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

2013-01-01

The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

Localization of the hot spots in a pebble bed reactor

International Nuclear Information System (INIS)

Chen, Leisheng; Lee, Wooram; Lee, Jaeyoung

2016-01-01

The pebble bed reactor (PBR) is a candidate reactor type for the very high temperature reactor (VHTR), which is one of the Generation-IV reactor types. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. The conclusions are made and may contribute to a better design of a PBR core and a closer inspection of the local hot spots to avoid destruction of pebbles from happening. Thermal field of a PBR core is investigated in this study. Specifically, experiments on measuring the pebbles' surface temperature are performed. It is found that the upper pebble has an overall higher temperature profile than the other pebbles and the stagnation zone under does not increase its surface's temperature. In addition, the temperature profile of the side pebble shows a concave form and it keeps decreasing from the contact point to the vertex in the lower pebble. Lastly, the maximum temperature difference among these points is 5.83 deg. C. These findings above are validated by CFX simulations under two different turbulence models (k-e, SST) and two contact areas (diameter of 6mm and 3.5mm). By contrasting the temperature variation trends of all simulation cases, it is concluded that SST turbulence model with 20% intensity shows a better agreement with the experiment result, nevertheless, slightly deviation is also found in terms of total temperature difference and the peak appears in position 17-19 in experiments

Localization of the hot spots in a pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

The pebble bed reactor (PBR) is a candidate reactor type for the very high temperature reactor (VHTR), which is one of the Generation-IV reactor types. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. The conclusions are made and may contribute to a better design of a PBR core and a closer inspection of the local hot spots to avoid destruction of pebbles from happening. Thermal field of a PBR core is investigated in this study. Specifically, experiments on measuring the pebbles' surface temperature are performed. It is found that the upper pebble has an overall higher temperature profile than the other pebbles and the stagnation zone under does not increase its surface's temperature. In addition, the temperature profile of the side pebble shows a concave form and it keeps decreasing from the contact point to the vertex in the lower pebble. Lastly, the maximum temperature difference among these points is 5.83 deg. C. These findings above are validated by CFX simulations under two different turbulence models (k-e, SST) and two contact areas (diameter of 6mm and 3.5mm). By contrasting the temperature variation trends of all simulation cases, it is concluded that SST turbulence model with 20% intensity shows a better agreement with the experiment result, nevertheless, slightly deviation is also found in terms of total temperature difference and the peak appears in position 17-19 in experiments.

Transient Effects in Fischer-Tropsch Reactor with a Fixed Bed of Catalyst Particles

Directory of Open Access Journals (Sweden)

I. V. Derevich

2015-01-01

Full Text Available Based on analysis of small temperature disturbances in the Fischer-Tropsch reactor with a fixed bed of catalyst particles various scenarios of thermal instability were investigated. There are two possible scenarios of thermal instability of the reactor. First, thermal explosion may occur due to growth of temperature disturbances inside a catalytic granule. Second scenario connected with loss of thermal stability as a result of an initial increase in temperature in the reactor volume. The boundaries of thermal stability of the reactor were estimated by solving the eigenvalue problems for spherical catalyst particles and cylindrical reactor. Processes of diffusional resistance inside the catalytic granule and heat transfer from wall of the reactor tube are taken into account. Estimation of thermal stability area is compared with the results of numerical simulation of behavior of temperature and concentration of synthesis gas.

Design and development of gas turbine high temperature reactor 300

International Nuclear Information System (INIS)

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

2003-01-01

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

Behavior of radioactive organic iodide in an atmosphere of High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Saeki, Masakatsu; Nakashima, Mikio; Sagawa, Chiaki; Masaki, Nobuyuki; Hirabayashi, Takakuni; Aratono, Yasuyuki

1990-06-01

Formation and decomposition behavior of radioactive organic iodide have been studied in an atmosphere of High Temperature Gas-cooled Reactor (High Temperature Engineering Test Reactor, HTTR). Na 125 I was chosen for radioactive iodine source instead of CsI diffusing from coated fuel particles. Na 125 I adsorbed on graphite was heated in pure He and He containing O 2 or H 2 O atmosphere. The results obtained are as follows. It was proved that organic iodide was formed with organic radicals released from graphite even in He atmosphere. Thus, the interchange rate of inorganic iodide with organic iodide was remarkably decreased with prolonged preheat-treatment period at 1000degC. Organic iodide formed was easily decomposed by its recirculation into hot reaction tube kept at 900degC. When organic iodide was passed through powdered graphite bed, more than 70% was decomposed at 90degC. Oxygen and water vapour intermixed in He suppressed the interchange rate of inorganic iodide with organic iodide. These results suggest that organic iodide rarely exists in the pressure vessel under normal operating condition of HTTR, and, under hypothetical accident condition of HTTR, organic iodide fraction never exceeds the value used for a safety assessment of light water reactor. (author)

The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 10

International Nuclear Information System (INIS)

Schulten, R.; Kugeler, K.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Woike, O.

1974-08-01

The necessary development steps, which have to be taken for the construction of a prototype plant for nuclear process heat, are enumerated. In particular, the work which is involved for the development of the nuclear steam-reforming technique, for the further development of the ball-shaped fuel elements at high gas outlet temperatures and for the reactor components, is described in detail. A brief survey of the needs of development of the IHX (intermediate heat exchanger) is given. An attempt is made to give overall time and cost estimates. (orig.) [de

High temperature filtration of radioactivable corrosion products in the primary circuit of PWR type reactors

International Nuclear Information System (INIS)

Dolle, L.

1976-01-01

A effective limitation to the deposition of radioactive corrosion products in the core of a reactor at power operation, is to be obtained by filtering the water of the primary circuit at a flow rate upper than 1% of the coolant flow rate. However, in view of accounting for more important release of corrosion products during the reactor start-up and also for some possible variations in the efficiency of the system, it is better that the flow rate to be treated by the cleaning circuit is stated at 5%. Filtration must be effected at the temperature of the primary circuit and preferably on each loop. To this end, the feasibility of electromagnetic filtration or filtration through a deep bed of granulated graphite has been studied. The on-loop tests effected on each filter gave efficiencies and yields respectively upper than 90% and 99% for magnetite and ferrite particles in suspension in water at 250 deg C. Such results confirm the interest lying in high temperature filtration and lead to envisage its application to reactors [fr

Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo [Korea Advanced Institute of Science and Tehcnology, Daejeon (Korea, Republic of)

2006-03-15

There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis.

Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

International Nuclear Information System (INIS)

Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo

2006-03-01

There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis

Hydrogen production from high temperature electrolysis and fusion reactor

International Nuclear Information System (INIS)

Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

1978-01-01

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

Pyrolysis of Softwood Carbohydrates in a Fluidized Bed Reactor

Directory of Open Access Journals (Sweden)

Dmitry Yu. Murzin

2008-09-01

Full Text Available In the present work pyrolysis of pure pine wood and softwood carbohydrates, namely cellulose and galactoglucomannan (the major hemicellulose in coniferous wood, was conducted in a batch mode operated fluidized bed reactor. Temperature ramping (5°C/min was applied to the heating until a reactor temperature of 460 °C was reached. Thereafter the temperature was kept until the release of non-condensable gases stopped. The different raw materials gave significantly different bio-oils. Levoglucosan was the dominant product in the cellulose pyrolysis oil. Acetic acid was found in the highest concentrations in both the galactoglucomannan and in the pine wood pyrolysis oils. Acetic acid is most likely formed by removal of O-acetyl groups from mannose units present in GGM structure.

Pebble bed modular reactor - The first Generation IV reactor to be constructed

International Nuclear Information System (INIS)

Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

2004-01-01

Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

Commissioning of the THTR-300-MWe prototype power plant - A milestone for further application of this high-temperature reactor line

International Nuclear Information System (INIS)

Simon, M.; Baust, E.; Schoening, J.

1986-10-01

With the completion of the THTR 300 and the development of the follow-on plant HTR 500, the BBC/HRB company group has taken the pebble bed high-temperature reactor to the threshold of the commercial stage. The HTR is an important innovation in the field of reactor technology which can play an important role in the intermediate and long-term supply of safe, environmental friendly and economic energy. The power level of 550 MW meets the requirements of the present energy market which shows a trend towards smaller power units as a result of grid size, investment effort, and the slower increase in electricity demand in industrial nations. The advantages of the high-temperature reactor, such as high thermal efficiency, low waste heat, low radiation exposure of operating and maintenance personnel, high inherent safety, simple mode of operation, flexible fuel cycle with the potential to extend fuel resources, high availability, are currently uncontested and will represent the future standards for the peaceful uses of nuclear energy. For special applications in industry (steam and electric power as a cogeneration product) and in case of special siting conditions (near industrial centers), BBC/HRB developed a small 100 MW HTR, which can also be constructed as a 200 MW twin plant at favorable cost conditions. For an economic use of domestic coal in a processed form, the HTR represents the optimum solution as to economic and environmental aspects as well as extension of resources, especially if combined with conventional gasification procedures and in direct application of nuclear process heat at high gas temperatures of about 950 deg. C. In this field the development of the heat-exchanging components remains to be completed, before commercial application will be possible. The HTR is particularly well suited for erection in developing countries and industrial threshold countries which turn to nuclear energy for the first time. On an international level the interest in the

Development of a trickle bed reactor of electro-Fenton process for wastewater treatment

International Nuclear Information System (INIS)

Lei, Yangming; Liu, Hong; Shen, Zhemin; Wang, Wenhua

2013-01-01

Highlights: • An electrochemical trickle bed reactor was composed of C-PTFE-coated graphite chips. • The trickle bed reactor had a high H 2 O 2 production rate in a dilute electrolyte. • An azo dye was effectively decomposed by the electro-Fenton process in the reactor. -- Abstract: To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H 2 O 2 was generated with a current of 0.3 A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte–cathode interface. In terms of H 2 O 2 generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L −1 of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3 h

Predicting freeboard heat transfer by using empirical correlations in high temperature fluidized beds

Energy Technology Data Exchange (ETDEWEB)

Biyikli, Suleyman [Okan University Tuzla Kampusu, Faculty of Engineering and Architecture (Turkey)], email: suleyman.biyikli@okan.edu.tr

2011-07-01

This article investigates the heat transfer characteristics for horizontal tubes in a freeboard region of high temperature fluidized beds. The freeboard entrainment heights are calculated by using empirical correlations described in detail and used in estimating the heat transfer coefficients from a horizontal tube occurring by radiation, gas convection, and particle contact mechanisms in high temperature a fluidized bed combustor. The total average of these coefficients around a horizontal tube carrying water in high temperature fluidized beds can be written as the sum of convective, radiative, and fluidized-particle contact heat transfer coefficients and these correlations are tested against certain published experimental measurements. In full agreement with this data, it was observed that the calculated heat transfer coefficients increased with increasing gas velocity at a given tube elevation and they decreased and approached the values of single-phase gas convection and radiation with increasing tube elevation in the freeboard region while the relative contribution of radiation increases and approaches a constant fraction of total heat transfer.

Sorption-enhanced steam methane reforming in fluidized bed reactors

Energy Technology Data Exchange (ETDEWEB)

Johnsen, Kim

2006-10-15

Hydrogen is considered to be an important potential energy carrier; however, its advantages are unlikely to be realized unless efficient means can be found to produce it without generation of CO{sub 2}. Sorption-enhanced steam methane reforming (SE-SMR) represent a novel, energy-efficient hydrogen production route with in situ CO{sub 2} capture, shifting the reforming and water gas shift reactions beyond their conventional thermodynamic limits. The use of fluidized bed reactors for SE-SMR has been investigated. Arctic dolomite, a calcium-based natural sorbent, was chosen as the primary CO{sub 2}-acceptor in this study due to high absorption capacity, relatively high reaction rate and low cost. An experimental investigation was conducted in a bubbling fluidized bed reactor of diameter 0.1 m, which was operated cyclically and batch wise, alternating between reforming/carbonation conditions and higher-temperature calcination conditions. Hydrogen concentrations of >98 mole% on a dry basis were reached at 600 C and 1 atm, for superficial gas velocities in the range of {approx}0.03-0.1 m/s. Multiple reforming-regeneration cycles showed that the hydrogen concentration remained at {approx}98 mole% after four cycles. The total production time was reduced with an increasing number of cycles due to loss of CO{sub 2}-uptake capacity of the dolomite, but the reaction rates of steam reforming and carbonation seemed to be unaffected for the conditions investigated. A modified shrinking core model was applied for deriving carbonation kinetics of Arctic dolomite, using experimental data from a novel thermo gravimetric reactor. An apparent activation energy of 32.6 kj/mole was found from parameter fitting, which is in good agreement with previous reported results. The derived rate expression was able to predict experimental conversion up to {approx}30% very well, whereas the prediction of higher conversion levels was poorer. However, the residence time of sorbent in a continuous

Engineering solutions for a reflector change concept in the high-temperature reactor with pebble bed core and OTTO-fueling

International Nuclear Information System (INIS)

Kasper, K.J.

1975-06-01

In the field of reactor engineering an increasing tendency is visible towards a 'repairable reactor'. In the construction of the HTR with spherical fuel elements this fact should already be taken into account at an early stage. Additionally it is possible that in connection with the OTTO-fueling load conditions for the graphite reflector could result which are locally not far away from limiting values. Therefore the removability of the reflector is included in the reactor construction as an accompanying technical step of the physical lay-out of the core. The core arrangements, realized for HTR until recently, are discussed as well as the properties of the graphites used and the operating conditions in the reactors are stated. At the example of the PR 3,000 proposals are offered for the construction of a removable side and top reflector for a pebble bed reactor. Hereby a solution was found which, on one hand allows the changing of the reflector and on the other hand requires no significant increase of the costs for the reactor assembly. Moreover the requirements of reactor operation and of repairability are satisfied in an optimal manner. (orig.) [de

Fluidized bed reactor for processing particles coated with carbon

International Nuclear Information System (INIS)

Marschollek, M.; Simon, W.; Walter, C.

1978-01-01

The carbon coating of production returns of these particles first has to be removed before the heavy metal core released can be reprocessed. For reasons of criticality, removal of burnt-up particles downwards must be possible in the fluidized bed reactor even if the reactor diameter is greater than 800 mm, and the material temperatures must not exceed 650 0 C. It consists of an upper cylindrical and a lower conical part, where, according to the invention, the gas distributor heads in the conical part are situated in several planes above one another for the fluidisation and combustion gas and where they are evently distributed over the reactor crossection, so that an even flow profile is achieved over the reactor cross section. (HP) [de

Core-adjacent instrumentation systems for pebble bed reactors for process heat application - state of planning

International Nuclear Information System (INIS)

Benninghofen, G.; Serafin, N.; Spillekothen, H.G.; Hecker, R.; Brixy, H.; Serpekian, T.

1982-06-01

Planning and theoretical/experimental development work for core surveillance instrumentation systems is being performed to meet requirements of pebble bed reactors for process heat application. Detailed and proved instrumentation concepts are now available for the core-adjacent instrumentation systems. The current work and the results of neutron flux measurements at high temperatures are described. Operation devices for long-term accurate gas outlet temperature measurements up to approximately 1423 deg. K will also be discussed. (author)

Optimization of the Neutronics of the Advanced High Temperature Reactor

International Nuclear Information System (INIS)

Zakova, Jitka; Talamo, Alberto

2006-01-01

In these studies, we have investigated the neutronic and safety performance of the Advanced High Temperature Reactor (AHTR) for plutonium and uranium fuels and we extended the analysis to five different coolants. The AHTR is a graphite-moderated and molten salt-cooled high temperature reactor, which takes advantage of the TRISO particles technology for the fuel utilization. The conceptual design of the core, proposed at the Oak Ridge National Laboratory, aims to provide an alternative to helium as coolant of high-temperature reactors for industrial applications like hydrogen production. We evaluated the influence of the radial reflector on the criticality of the core for the uranium and plutonium fuels and we focused on the void coefficient of 5 different molten salts; since the safety of the reactor is enhanced also by the large and negative coefficient of temperature, we completed our investigation by observing the keff changes when the graphite temperature varies from 300 to 1800 K. (authors)

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Ultra-high temperature direct propulsion

International Nuclear Information System (INIS)

Araj, K.J.; Slovik, G.; Powell, J.R.; Ludewig, H.

1987-01-01

Potential advantages of ultra-high exhaust temperature (3000 K - 4000 K) direct propulsion nuclear rockets are explored. Modifications to the Particle Bed Reactor (PBR) to achieve these temperatures are described. Benefits of ultra-high temperature propulsion are discussed for two missions - orbit transfer (ΔV = 5546 m/s) and interplanetary exploration (ΔV = 20000 m/s). For such missions ultra-high temperatures appear to be worth the additional complexity. Thrust levels are reduced substantially for a given power level, due to the higher enthalpy caused by partial disassociation of the hydrogen propellant. Though technically challenging, it appears potentially feasible to achieve such ultra high temperatures using the PBR

Creep behavior of materials for high-temperature reactor application

International Nuclear Information System (INIS)

Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

1984-01-01

Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

Anaerobic treatment of winery wastewater in fixed bed reactors.

Science.gov (United States)

Ganesh, Rangaraj; Rajinikanth, Rajagopal; Thanikal, Joseph V; Ramanujam, Ramamoorty Alwar; Torrijos, Michel

2010-06-01

The treatment of winery wastewater in three upflow anaerobic fixed-bed reactors (S9, S30 and S40) with low density floating supports of varying size and specific surface area was investigated. A maximum OLR of 42 g/l day with 80 +/- 0.5% removal efficiency was attained in S9, which had supports with the highest specific surface area. It was found that the efficiency of the reactors increased with decrease in size and increase in specific surface area of the support media. Total biomass accumulation in the reactors was also found to vary as a function of specific surface area and size of the support medium. The Stover-Kincannon kinetic model predicted satisfactorily the performance of the reactors. The maximum removal rate constant (U(max)) was 161.3, 99.0 and 77.5 g/l day and the saturation value constant (K(B)) was 162.0, 99.5 and 78.0 g/l day for S9, S30 and S40, respectively. Due to their higher biomass retention potential, the supports used in this study offer great promise as media in anaerobic fixed bed reactors. Anaerobic fixed-bed reactors with these supports can be applied as high-rate systems for the treatment of large volumes of wastewaters typically containing readily biodegradable organics, such as the winery wastewater.

Thermofluid-neutronic stability of the rotating, fluidized bed, space-power reactor

International Nuclear Information System (INIS)

Lee, C.C.; Jones, O.C.; Becker, M.

1993-01-01

A rotating fluidized bed nuclear reactor has the potential of being a vary attractive option for ultra-high power space systems, especially for propulsion. Research has already examined fuel bed expansion due to variations in state variables, propellant flow rate, and rotational speed, and has also considered problems related to thermal stress. This paper describes the results of a coupled thermofluid-neutronic analysis where perturbations in fuel bed height caused by maneuvering changes in operating conditions alter power levels due to varying absorption of neutrons which would otherwise leak from the system, mainly through the nozzle. This first analysis was not a detailed stability analysis. Rather, it utilized simplified neutronic methods, and was intended to provide an order-of-magnitude assessment of the stability of the reactor with the intention to determine whether or not stability might be a 'concept killer'. Stability was compared with a fixed-fuel-bed reactor of identical geometry for three different cases comprising a set of small, medium and large sizes/powers from 250 MW to 5 GW. It was found that power fluctuations in the fluidized bed reactor were larger by 100 db or more than expected in a packed bed reactor of the same geometry, but never resulted in power excursions. Margins to unit gain in some cases, however, were sufficiently small that the approximations in this quasi-2-dimensional model may not be sufficiently accurate to preclude significant excursions. (orig.)

High temperature degradation by erosion-corrosion in bubbling fluidized bed combustors

Directory of Open Access Journals (Sweden)

Hou Peggy

2004-01-01

Full Text Available Heat-exchanger tubes in fluidized bed combustors (FBCs often suffer material loss due to combined corrosion and erosion. Most severe damage is believed to be caused by the impact of dense packets of bed material on the lower parts of the tubes. In order to understand this phenomenon, a unique laboratory test rig at Berkeley was designed to simulate the particle hammering interactions between in-bed particles and tubes in bubbling fluidized bed combustors. In this design, a rod shaped specimen is actuated a short distance within a partially fluidized bed. The downward specimen motion is controlled to produce similar frequencies, velocities and impact forces as those experienced by the impacting particle aggregates in practical systems. Room temperature studies have shown that the degradation mechanism is a three-body abrasion process. This paper describes the characteristics of this test rig, reviews results at elevated temperatures and compares them to field experience. At higher temperatures, deposits of the bed material on tube surfaces can act as a protective layer. The deposition depended strongly on the type of bed material, the degree of tube surface oxidation and the tube and bed temperatures. With HCl present in the bed, wastage was increased due to enhanced oxidation and reduced oxide scale adherence.

Pebble bed reactor fuel cycle optimization using particle swarm algorithm

Energy Technology Data Exchange (ETDEWEB)

Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

2016-10-15

Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

Pebble bed reactor fuel cycle optimization using particle swarm algorithm

International Nuclear Information System (INIS)

Tavron, Barak; Shwageraus, Eugene

2016-01-01

Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

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

International Nuclear Information System (INIS)

Sudo, Yukio

1990-01-01

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

An in situ spatially resolved analytical technique to simultaneously probe gas phase reactions and temperature within the packed bed of a plug flow reactor.

Science.gov (United States)

Touitou, Jamal; Burch, Robbie; Hardacre, Christopher; McManus, Colin; Morgan, Kevin; Sá, Jacinto; Goguet, Alexandre

2013-05-21

This paper reports the detailed description and validation of a fully automated, computer controlled analytical method to spatially probe the gas composition and thermal characteristics in packed bed systems. As an exemplar, we have examined a heterogeneously catalysed gas phase reaction within the bed of a powdered oxide supported metal catalyst. The design of the gas sampling and the temperature recording systems are disclosed. A stationary capillary with holes drilled in its wall and a moveable reactor coupled with a mass spectrometer are used to enable sampling and analysis. This method has been designed to limit the invasiveness of the probe on the reactor by using the smallest combination of thermocouple and capillary which can be employed practically. An 80 μm (O.D.) thermocouple has been inserted in a 250 μm (O.D.) capillary. The thermocouple is aligned with the sampling holes to enable both the gas composition and temperature profiles to be simultaneously measured at equivalent spatially resolved positions. This analysis technique has been validated by studying CO oxidation over a 1% Pt/Al2O3 catalyst and the spatial resolution profiles of chemical species concentrations and temperature as a function of the axial position within the catalyst bed are reported.

Optimization of a radially cooled pebble bed reactor - HTR2008-58117

International Nuclear Information System (INIS)

Boer, B.; Kloosterman, J. L.; Lathouwers, D.; Van Der Hagen, T. H. J. J.; Van Dam, H.

2008-01-01

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as I cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = -2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = -50 deg. C) can be achieved

Development of a trickle bed reactor of electro-Fenton process for wastewater treatment

Energy Technology Data Exchange (ETDEWEB)

Lei, Yangming [Key Laboratory of Reservoir Aquatic Environment, Chongqing Institute of Green and Intelligent Technology, Chinese Academy of Sciences, Chongqing 401122 (China); School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Liu, Hong, E-mail: liuhong@cigit.ac.cn [Key Laboratory of Reservoir Aquatic Environment, Chongqing Institute of Green and Intelligent Technology, Chinese Academy of Sciences, Chongqing 401122 (China); Shen, Zhemin, E-mail: zmshen@sjtu.edu.cn [School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Wang, Wenhua [School of Environmental Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China)

2013-10-15

Highlights: • An electrochemical trickle bed reactor was composed of C-PTFE-coated graphite chips. • The trickle bed reactor had a high H{sub 2}O{sub 2} production rate in a dilute electrolyte. • An azo dye was effectively decomposed by the electro-Fenton process in the reactor. -- Abstract: To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H{sub 2}O{sub 2} was generated with a current of 0.3 A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte–cathode interface. In terms of H{sub 2}O{sub 2} generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L{sup −1} of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3 h.

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.

Experiment and modeling of low-concentration methane catalytic combustion in a fluidized bed reactor

International Nuclear Information System (INIS)

Yang, Zhongqing; Yang, Peng; Zhang, Li; Guo, Mingnv; Ran, Jingyu

2016-01-01

Highlights: • The catalytic combustion of 0.15~3 vol. % low concentration methane in a fluidized bed was studied. • A mathematical model was proposed on the basis of gas–solid flow theory. • A comparative analysis of the established model with plug flow, mixed flow and K-L models was carried out. • The axial methane profile along fluidized bed was predicted by using the mathematical model. • The bed temperature has greater impact on methane conversion than fluidized velocity. - Abstract: This study undertakes a theoretical analysis and an experimental investigation into the characteristics of low-concentration methane catalytic combustion in a bubbling fluidized bed reactor using 0.5 wt.% Pd/Al_2O_3 as catalytic particles. A mathematical model is established based on gas–solid flow theory and is used to study the effects of bed temperature and fluidized velocity on methane catalytic combustion, and predict the dimensionless methane concentration axial profile in reactor. It is shown that methane conversion increases with bed temperature, but decreases with increasing fluidized velocity. These theoretical results are found to correlate well with the experimental measurement, with a deviation within 5%. A comparative analysis of the developed model with plug flow, mixed flow and K-L models is also carried out, and this further verifies that the established model better reflects the characteristics of low-concentration methane catalytic combustion in a bubbling fluidized bed. Using this reaction model, it was found that the difference in methane conversion between dense and freeboard zones gradually increases with bed temperature; the dense zone reaction levels off at 650 °C, thereby minimizing the difference between the dense and freeboard regions to around 15%. With an increase in bed temperature, the dimensionless methane concentration in the dense zone decreases exponentially, while in the splash zone, it varies from an exponential decay to a slow

Anaerobic up flow fluidized bed reactor performance as a primary treatment unit in domestic wastewater treatment

Directory of Open Access Journals (Sweden)

M.A. Moharram

2016-04-01

The efficiencies of Total nitrogen removal ranged between 2.23 and 10.83% with an apparent decrease during the low temperature high rate stages. Nitrite removal was in the range of (23.08–77% with up to the 2 mg/L in the effluent water when obtaining high organic loading and warm temperature. These results demonstrated that the domestic wastewater could be anaerobically treated in a fluidized bed UASB reactor with very low HRT reaching 2.5 h.

Conversion of metallurgical coke and coal using a Coal Direct Chemical Looping (CDCL) moving bed reactor

International Nuclear Information System (INIS)

Luo, Siwei; Bayham, Samuel; Zeng, Liang; McGiveron, Omar; Chung, Elena; Majumder, Ankita; Fan, Liang-Shih

2014-01-01

Highlights: • Accumulated more than 300 operation hours were accomplished for the moving bed reducer reactor. • Different reactor operation variables were investigated with optimal conditions identified. • High conversions of sub-bituminous coal and bituminous coal were achieved without flow problems. • Co-current and counter-current contact modes were tested and their applicability was discussed. - Abstract: The CLC process has the potential to be a transformative commercial technology for a carbon-constrained economy. The Ohio State University Coal Direct Chemical Looping (CDCL) process directly converts coal, eliminating the need for a coal gasifier oran air separation unit (ASU). Compared to other solid-fuel CLC processes, the CDCL process is unique in that it consists of a countercurrent moving bed reducer reactor. In the proposed process, coal is injected into the middle of the moving bed, whereby the coal quickly heats up and devolatilizes, splitting the reactor roughly into two sections with no axial mixing. The top section consists of gaseous fuel produced from the coal volatiles, and the bottom section consists of the coal char mixed with the oxygen carrier. A bench-scale moving bed reactor was used to study the coal conversion with CO 2 as the enhancing gas. Initial tests using metallurgical cokefines as feedstock were conducted to test the effects of operational variables in the bottom section of the moving bed reducer, e.g., reactor temperature, oxygen carrier to char ratio, enhancer gas CO 2 flow rate, and oxygen carrier flow rates. Experiments directly using coal as the feedstock were subsequently carried out based on these test results. Powder River Basin (PRB) coal and Illinois #6 coal were tested as representative sub-bituminous and bituminous coals, respectively. Nearly complete coal conversion was achieved using composite iron oxide particles as the oxygen carriers without any flow problems. The operational results demonstrated that a

Computational fluid dynamic modeling of fluidized-bed polymerization reactors

Energy Technology Data Exchange (ETDEWEB)

Rokkam, Ram [Iowa State Univ., Ames, IA (United States)

2012-01-01

Polyethylene is one of the most widely used plastics, and over 60 million tons are produced worldwide every year. Polyethylene is obtained by the catalytic polymerization of ethylene in gas and liquid phase reactors. The gas phase processes are more advantageous, and use fluidized-bed reactors for production of polyethylene. Since they operate so close to the melting point of the polymer, agglomeration is an operational concern in all slurry and gas polymerization processes. Electrostatics and hot spot formation are the main factors that contribute to agglomeration in gas-phase processes. Electrostatic charges in gas phase polymerization fluidized bed reactors are known to influence the bed hydrodynamics, particle elutriation, bubble size, bubble shape etc. Accumulation of electrostatic charges in the fluidized-bed can lead to operational issues. In this work a first-principles electrostatic model is developed and coupled with a multi-fluid computational fluid dynamic (CFD) model to understand the effect of electrostatics on the dynamics of a fluidized-bed. The multi-fluid CFD model for gas-particle flow is based on the kinetic theory of granular flows closures. The electrostatic model is developed based on a fixed, size-dependent charge for each type of particle (catalyst, polymer, polymer fines) phase. The combined CFD model is first verified using simple test cases, validated with experiments and applied to a pilot-scale polymerization fluidized-bed reactor. The CFD model reproduced qualitative trends in particle segregation and entrainment due to electrostatic charges observed in experiments. For the scale up of fluidized bed reactor, filtered models are developed and implemented on pilot scale reactor.

Characteristic Studies of Micron Zinc Particle Hydrolysis in a Fixed Bed Reactor

Directory of Open Access Journals (Sweden)

Lv Ming

2015-09-01

Full Text Available Zinc fuel is considered as a kind of promising energy sources for marine propeller. As one of the key steps for zinc marine energy power system, zinc hydrolysis process had been studied experimentally in a fixed bed reactor. In this study, we focus on the characteristics of micron zinc particle hydrolysis. The experimental results suggested that the steam inner diffusion is the controlling step of accumulative zinc particles hydrolysis reaction at a relative lower temperature and a relative higher water partial pressure. In other conditions, the chemical reaction kinetics was the controlling step. And two kinds of chemical reaction kinetics appeared in experiments: the surface reaction and the gas-gas reaction. The latter one occurs usually for larger zinc particles and high reaction temperature. Temperature seems to be one of the most important parameters for the dividing of different reaction mechanisms. Several parameters of the hydrolysis process including heating rate, water partial pressure, the particle size and temperature were also studied in this paper. Results show that the initial reaction temperature of zinc hydrolysis in fixed bed is about 410°C. And the initial reaction temperature increases as the heating rate increases and as the water partial pressure decreases. The total hydrogen yield increases as the heating rate decreases, as the water partial pressure increases, as the zinc particle size decreases, and as the reaction temperature increases. A hydrogen yield of more than 81.5% was obtained in the fixed bed experiments.

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

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

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.)

Accounting for porous structure in effective thermal conductivity calculations in a pebble bed reactor

International Nuclear Information System (INIS)

Antwerpen, W. van; Rousseau, P.G.; Toit, C.G. du

2009-01-01

A proper understanding of the mechanisms of heat transfer, flow and pressure drop through a packed bed of spheres is of utmost importance in the design of a high temperature pebble bed reactor. A thorough knowledge of the porous structure within the packed bed is important to any rigorous analysis of the transport phenomena, as all the heat and flow mechanisms are influenced by the porous structure. In this paper a new approach is proposed to simulate the effective thermal conductivity employing a combination of new and existing correlations for randomly packed beds. More attention is given to packing structure based on coordination number and contact angles, resulting in a more rigorous differentiation between the bulk and near-wall regions. The model accounts for solid conduction, gas conduction, contact area, surface roughness as well as radiation. (author)

Continuous biological waste gas treatment in stirred trickle-bed reactor with discontinuous removal of biomass.

Science.gov (United States)

Laurenzis, A; Heits, H; Wübker, S; Heinze, U; Friedrich, C; Werner, U

1998-02-20

A new reactor for biological waste gas treatment was developed to eliminate continuous solvents from waste gases. A trickle-bed reactor was chosen with discontinuous movement of the packed bed and intermittent percolation. The reactor was operated with toluene as the solvent and an optimum average biomass concentration of between 5 and 30 kg dry cell weight per cubic meter packed bed (m3pb). This biomass concentration resulted in a high volumetric degradation rate. Reduction of surplus biomass by stirring and trickling caused a prolonged service life and prevented clogging of the trickle bed and a pressure drop increase. The pressure drop after biomass reduction was almost identical to the theoretical pressure drop as calculated for the irregular packed bed without biomass. The reduction in biomass and intermittent percolation of mineral medium resulted in high volumetric degradation rates of about 100 g of toluene m-3pb h-1 at a load of 150 g of toluene m-3pb h-1. Such a removal rate with a trickle-bed reactor was not reported before. Copyright 1998 John Wiley & Sons, Inc.

Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

International Nuclear Information System (INIS)

Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

2006-01-01

Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

2013-01-01

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

2013-07-15

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

Systems and methods for enhancing isolation of high-temperature reactor containments

Energy Technology Data Exchange (ETDEWEB)

Peterson, Per F.

2017-09-26

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

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

International Nuclear Information System (INIS)

Forsberg, Charles W.

2001-01-01

Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

Directory of Open Access Journals (Sweden)

L.P. Rodriguez

2015-08-01

Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

Summary of ORNL high-temperature gas-cooled reactor program

International Nuclear Information System (INIS)

Kasten, P.R.

1981-01-01

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

Study on thermodynamic cycle of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu Xinhe; Yang Xiaoyong; Wang Jie

2017-01-01

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

Analysis of the start-up and control of a particle bed reactor

International Nuclear Information System (INIS)

Lazareth, O.W.; Araj, K.J.; Horn, F.L.; Ludewig, H.; Powell, J.R.

1987-01-01

This study describes the modeling of start-up transients in Particle Bed Reactors (PBR) for burst electric power. Two computer programs have been developed to analyze the start-up process. The first program (named KINETIC) analyzes the entire fuel element, calculating time dependent solutions for power and the temperature distribution in the packed bed. The second program (named SPHEAT, for Spherical Heating) calculates time-dependent temperatures inside individual, cladded fuel particles. The two programs provide powerful analytical tools for evaluation of material and geometrical options, power and time constraints, and conditions that could lead to element failures

Bubbling bed catalytic hydropyrolysis process utilizing larger catalyst particles and smaller biomass particles featuring an anti-slugging reactor

Science.gov (United States)

Marker, Terry L; Felix, Larry G; Linck, Martin B; Roberts, Michael J

2014-09-23

This invention relates to a process for thermochemically transforming biomass or other oxygenated feedstocks into high quality liquid hydrocarbon fuels. In particular, a catalytic hydropyrolysis reactor, containing a deep bed of fluidized catalyst particles is utilized to accept particles of biomass or other oxygenated feedstocks that are significantly smaller than the particles of catalyst in the fluidized bed. The reactor features an insert or other structure disposed within the reactor vessel that inhibits slugging of the bed and thereby minimizes attrition of the catalyst. Within the bed, the biomass feedstock is converted into a vapor-phase product, containing hydrocarbon molecules and other process vapors, and an entrained solid char product, which is separated from the vapor stream after the vapor stream has been exhausted from the top of the reactor. When the product vapor stream is cooled to ambient temperatures, a significant proportion of the hydrocarbons in the product vapor stream can be recovered as a liquid stream of hydrophobic hydrocarbons, with properties consistent with those of gasoline, kerosene, and diesel fuel. Separate streams of gasoline, kerosene, and diesel fuel may also be obtained, either via selective condensation of each type of fuel, or via later distillation of the combined hydrocarbon liquid.

Bubbling bed catalytic hydropyrolysis process utilizinig larger catalyst particles and small biomass particles featuring an anti-slugging reactor

Science.gov (United States)

Marker, Terry L.; Felix, Larry G.; Linck, Martin B.; Roberts, Michael J.

2016-12-06

This invention relates to a process for thermochemically transforming biomass or other oxygenated feedstocks into high quality liquid hydrocarbon fuels. In particular, a catalytic hydropyrolysis reactor, containing a deep bed of fluidized catalyst particles is utilized to accept particles of biomass or other oxygenated feedstocks that are significantly smaller than the particles of catalyst in the fluidized bed. The reactor features an insert or other structure disposed within the reactor vessel that inhibits slugging of the bed and thereby minimizes attrition of the catalyst. Within the bed, the biomass feedstock is converted into a vapor-phase product, containing hydrocarbon molecules and other process vapors, and an entrained solid char product, which is separated from the vapor stream after the vapor stream has been exhausted from the top of the reactor. When the product vapor stream is cooled to ambient temperatures, a significant proportion of the hydrocarbons in the product vapor stream can be recovered as a liquid stream of hydrophobic hydrocarbons, with properties consistent with those of gasoline, kerosene, and diesel fuel. Separate streams of gasoline, kerosene, and diesel fuel may also be obtained, either via selective condensation of each type of fuel, or via later distillation of the combined hydrocarbon liquid.

Effect of temperature, hydraulic residence time and elevated PCO2 on acid neutralization within a pulsed limestone bed reactor

Science.gov (United States)

Watten, B.J.; Lee, P.C.; Sibrell, P.L.; Timmons, M.B.

2007-01-01

Limestone has potential for reducing reagent costs and sludge volume associated with treatment of acid mine drainage, but its use is restricted by slow dissolution rates and the deposition of Fe, Al and Mn-based hydrolysis products on reactive surfaces. We evaluated a pulsed limestone bed (PLB) reactor (15 L/min capacity) that uses a CO2 pretreatment step to accelerate dissolution and hydraulic shearing forces provided by intermittent fluidization to abrade and carry away surface scales. We established the effects of hydraulic residence time (HRT, 5.1-15.9 min), temperature (T, 12-22 ??C) and CO2 tension (PCO2, 34.5-206.8 kPa) on effluent quality when inlet acidity (Acy) was fixed at 440 mg/L (pH=2.48) with H2SO4. The PLB reactor neutralized all H+ acidity (N=80) while concurrently providing unusually high levels of effluent alkalinity (247-1028 mg/L as CaCO3) that allow for side-stream treatment with blending. Alkalinity (Alk) yields rose with increases in PCO2, HRT and settled bed height (BH, cm) and decreased with T following the relationship (R2=0.926; p<0.001): (Alk)non-filtered=-548.726+33.571??(PCO2)0.5+33.671??(HRT)+7.734??(BH)-5.197??(T). Numerical modeling showed CO2 feed requirements for a target Alk yield decrease with increases in HRT, T and the efficiency of off-gas (CO2) recycling. ?? 2007 Elsevier Ltd. All rights reserved.

The Integration Of Process Heat Applications To High Temperature Gas Reactors

International Nuclear Information System (INIS)

McKellar, Michael G.

2011-01-01

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

A preliminary neutronic evaluation of the high temperature gas-cooled test reactor HTR-10 using the scale 6.0 code

International Nuclear Information System (INIS)

Sousa, Romulo V.; Fortini, Angela; Pereira, Claubia; Carvalho, Fernando R. de; Oliveira, Arno H.

2013-01-01

The High Temperature Gas-cooled Test Reactor HTR-10 is a 10 MW modular pebble bed type reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles. This reactor was built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China, and its first criticality was attained on December 1, 2000. The main objectives of the HTR-10 are to verify and demonstrate the technical and safety features of the modular HTGR (High Temperature Gas-cooled Reactor) and to establish an experimental base for developing nuclear process heat applications. In this work, using the Standardized Computer Analysis for Licensing Evaluation (SCALE) 6.0, a nuclear code developed by Oak Ridge National Laboratory (ORNL), the HTR-10 first critical core is modeled by the DEN/UFMG. The K eff was obtained and compared with the reference value obtained by the Idaho National Laboratory. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel. (author)

Effect of a flow-corrective insert on the flow pattern in a pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Li, Yu; Gui, Nan; Yang, Xingtuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Tu, Jiyuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); School of Aerospace, Mechanical & Manufacturing Engineering, RMIT University, Melbourne 3083, VIC (Australia); Jiang, Shengyao, E-mail: shengyaojiang@sina.com [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-04-15

Highlights: • Effect of an insert on improving flow uniformity and eliminating stagnant zone is studied. • Three values concerned with the stagnant zone, radial uniformity and flow sequence are used. • Outlet diameter is a critical parameter that determines balancing mechanism of the insert. • Height/location is varied to let the insert work in unbalanced region and avoid adverse effect. - Abstract: A flow-corrective insert is adopted in the pebble-bed high temperature gas-cooled reactor (HTGR) to improve flow performance of the pebble flow for the first time. 3D discrete element method (DEM) modeling is employed to study this slow and dense granular flow. It is verified that locating a properly designed insert in the bed can help transform unsatisfactory flow field to the preferred flow pattern for pebble bed reactors. Three characteristic values on the stagnant zone, radial uniformity and flow sequence of pebble flow are defined to evaluate uniformity of the overall flow field quantitatively. The results demonstrate that the pebble bed equipped with an insert performs better than normal beds from all these three aspects. Moreover, based on numerical experiments, several universal tips for insert design on height, location and outlet diameter are suggested.

Study of the obtainment of Mo_2C by gas-solid reaction in a fixed and rotary bed reactor

International Nuclear Information System (INIS)

Araujo, C.P.B. de; Souza, C.P. de; Souto, M.V.M.; Barbosa, C.M.; Frota, A.V.V.M.

2016-01-01

Carbides' synthesis via gas-solid reaction overcomes many of the difficulties found in other processes, requiring lower temperatures and reaction times than traditional metallurgic routes, for example. In carbides' synthesis in fixed bed reactors (FB) the solid precursor is permeated by the reducing/carburizing gas stream forming a packed bed without mobility. The use of a rotary kiln reactor (RK) adds a mixing character to this process, changing its fluid-particle dynamics. In this work ammonium molybdate was subjected to carbo-reduction reaction (CH4 / H2) in both reactors under the same gas flow (15L / h) and temperature (660 ° C) for 180 minutes. Complete conversion was observed Mo2C (dp = 18.9nm modal particles sizes' distribution) in the fixed bed reactor. In the RK reactor this conversion was only partial (∼ 40%) and Mo2C and MoO3 (34nm dp = bimodal) could be observed on the produced XRD pattern. Partial conversion was attributed to the need to use higher solids loading in the reactor CR (50% higher) to avoid solids to centrifuge. (author)

Studies on high temperature research reactor in China

Energy Technology Data Exchange (ETDEWEB)

Xu Yuanhui; Zuo Kanfen [Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)

1999-08-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

Studies on high temperature research reactor in China

International Nuclear Information System (INIS)

Xu Yuanhui; Zuo Kanfen

1999-01-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

CFD analysis of hot spot formation through a fixed bed reactor of Fischer-Tropsch synthesis

Directory of Open Access Journals (Sweden)

Hamed Aligolzadeh

2015-12-01

Full Text Available One of the interesting methods for conversion of synthesis gas to heavy hydrocarbons is Fischer–Tropsch process. The process has some bottlenecks, such as hot spot formation and low degree of conversion. In this work, computational fluid dynamics technique was used to simulate conversion of synthetic gas and product distribution. Also, hot spot formation in the catalytic fixed-bed reactor was investigated in several runs. Simulation results indicated that hot spot formation occurred more likely in the early and middle part of reactor due to high reaction rates. Based on the simulation results, the temperature of hot spots increased with increase in the inlet temperature as well as pressure. Among the many CFD runs conducted, it is found that the optimal temperature and pressure for Fischer–Tropsch synthesis are 565 K and 20 bar, respectively. As it seems that the reactor shall work very well under optimal conditions, the reaction rates and catalyst duration would simultaneously be maximum .

INTENSIFICATION OF HEAT TRANSFER IN A HIGH-TEMPERATURED FLUIDIZED BED

Directory of Open Access Journals (Sweden)

А. O. Redko

2017-10-01

Full Text Available Purpose. This paper highlights experimental research of heat exchange in coarse particles to ensure the performance of ecological characteristics of Heat supply system. Methodology. The test stand has been developed to solve the defined task. It helps to do the research at the temperature of fluidizing bed and pulsating fluidizing bed at the range 800–1000°С. The temperature of the fluidized bed was provided by burning natural gas and wood waste. Sand and chamotte with a particle size of 1.0 to 5.0 mm were used as the layer material. The heat-transfer coefficient from the layer to the surface, immersed in the layer, and the density of the heat flux were measured by a calorimetric method under steady-state conditions. Smooth tubes and transversely finned with different height and rib spacing were investigated. Experiments in a high-temperature pulsating fluidized bed were carried out with pulsating combustion of natural gas in a layer or in a sublattice chamber into which natural gas and air were separately supplied. The frequency of combustion pulsations was provided by the automation system. The flash frequency was regulated in the range from 0.14 to 5 Hz. Findings. It is presented the results of physical modeling to find out the heat-transfer coefficients of smooth and finned tubes in fluidizing bed of coarse particles in the process of wood waste and gaseous fuel combusting. It is proved that the coefficient of heat transfer increases with increasing temperature by 2–2,5 times in the bed which contains particles diameter of 2.5–5 mm is 300–350Vt/(m2 К that is much higher than for layer furnaces. The results of the experiments are presented in the form of a generalized relationship that takes into account the diameter of the particles and the value of the finning coefficient. Heat transfer of finned tubes is 15–20% less then smooth tubes but the density of heat flow referred to the area of a finned tube is 0,12–0,20 МVt/m2 that

Particle bed reactor modeling

Science.gov (United States)

Sapyta, Joe; Reid, Hank; Walton, Lew

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

High-temperature reactor in modular construction

International Nuclear Information System (INIS)

Mueller, F.U.; Reutler, H.; Ullrich, M.

1981-01-01

Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

The primary circuit of the dragon high temperature reactor experiment

International Nuclear Information System (INIS)

Simon, R.

2005-01-01

The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

Fungi solubilisation of low rank coal: performances of stirred tank, fluidised bed and packed bed reactors

CSIR Research Space (South Africa)

Oboirien, BO

2013-02-01

Full Text Available Coal biosolubilisation was investigated in stirred tank reactor, fluidised bed and fixed bed bioreactors with a view to highlight the advantages and shortcomings of each of these reactor configurations. The stirred aerated bioreactor and fluidised...

Coupling membrane pervaporation with a fixed-bed reactor for enhanced esterification of oleic acid with ethanol

International Nuclear Information System (INIS)

Han, Ying; Lv, Enmin; Ma, Lingling; Lu, Jie; Chen, Kexun; Ding, Jincheng

2015-01-01

Highlights: • The reactor coupling membrane pervaporation with a fixed-bed reactor was studied. • The factors effecting the esterification of oleic acid were investigated. • NaA zeolite membrane was used for dehydration in the coupled reactor. - Abstract: Process intensification through membrane pervaporation (PV) integrated with a fixed-bed reactor could be successfully applied to the esterification of oleic acid and ethanol, which is a crucial step in the biodiesel synthesis using waste oil and grease as resource. The properties of the NaA zeolite membrane such as structure, formulation and separation were investigated by scanning electronic microscopy–energy dispersive spectrometry (SEM–EDS), X-ray diffractometry (XRD) and PV dehydration. Results showed that the NaA zeolite membrane had good separating property for removing water from the organics mixture. The operating conditions were optimized as the ethanol to oleic acid molar ratio of 15:1, feedstock flow rate of 1.0 ml/min, reaction temperature of 80.0 °C and catalyst bed height of 132 mm. The final conversion of oleic acid increased from 84.23% to 87.18% by PV using the NaA zeolite membrane at 24.0 h of operation. The membrane showed good PV performance after used for eight successive runs in the PV-assisted esterification. The resin exhibited a much high catalytic activity and operation stability after used for 100 h in the consecutive single pass fixed-bed esterification.

Effects of water vapor pretreatment time and reaction temperature on CO(2) capture characteristics of a sodium-based solid sorbent in a bubbling fluidized-bed reactor.

Science.gov (United States)

Seo, Yongwon; Jo, Sung-Ho; Ryu, Chong Kul; Yi, Chang-Keun

2007-10-01

CO(2) capture from flue gas using a sodium-based solid sorbent was investigated in a bubbling fluidized-bed reactor. Carbonation and regeneration temperature on CO(2) removal was determined. The extent of the chemical reactivity after carbonation or regeneration was characterized via (13)C NMR. In addition, the physical properties of the sorbent such as pore size, pore volume, and surface area after carbonation or regeneration were measured by gas adsorption method (BET). With water vapor pretreatment, near complete CO(2) removal was initially achieved and maintained for about 1-2min at 50 degrees C with 2s gas residence time, while without proper water vapor pretreatment CO(2) removal abruptly decreased from the beginning. Carbonation was effective at the lower temperature over the 50-70 degrees C temperature range, while regeneration more effective at the higher temperature over the 135-300 degrees C temperature range. To maintain the initial 90% CO(2) removal, it would be necessary to keep the regeneration temperature higher than about 135 degrees C. The results obtained in this study can be used as basic data for designing and operating a large scale CO(2) capture process with two fluidized-bed reactors.

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

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

Mechanical properties of concrete for power reactor at high temperatures

International Nuclear Information System (INIS)

Kawase, Kiyotaka; Tanaka, Hitoshi; Nakano, Masayuki

1985-01-01

The purpose of this study is to investigate the mechanical properties of concrete for power reactor at high temperature. This paper presents the creep behavior of concrete at high temperature and the cause by which a specified aggregate is broken at a specified high temperature. The creep coefficient at high temperature is smaller than that at ordinary temperature. (author)

New temperature monitoring devices for high-temperature irradiation experiments in the high flux reactor Petten

Energy Technology Data Exchange (ETDEWEB)

Laurie, M.; Futterer, M. A.; Lapetite, J. M. [European Commission Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, 61438 Flers Cedex (France); Morice, R. [Laboratoire National de Metrologie et d' Essais, 1 rue Gaston Boissier, 75724 Paris (France)

2009-07-01

Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some research and development work for further improvement of thermocouples and other on-line instrumentation will be outlined. (authors)

Mathematical modelling of fluidized bed reactors

Energy Technology Data Exchange (ETDEWEB)

Werther, J [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1978-11-01

Among the many fluidized bed models to be found in the literature, the two-phase model originally proposed by May has proved most suitable for accomodation of recent advances in flow mechanics: this model resolves the gas/solids fluidized bed into a bubble phase and a suspension phase surrounding the bubbles. Its limitation to slow reactions is a disadvantage. On the basis of the analogy between fluidized beds and gas/liquid systems, a general two-phase model that is valid for fast reactions has therefore been developed and its validity is confirmed by comparison with the experimental results obtained by others. The model describes mass transfer across the phase interface with the aid of the film theory known from gas/liquid reactor technology, and the reaction occurring in the suspension phase as a pseudo-homogeneous reaction. Since the dependence of the performance of fluidized bed reactors upon geometry is accounted for, the model can also be used for scale-up calculations. Its use is illustrated with the aid of design diagrams.

Fast pyrolysis of Miscanthus sinensis in fluidized bed reactors: Characteristics of product yields and biocrude oil quality

International Nuclear Information System (INIS)

Bok, Jin Pil; Choi, Hang Seok; Choi, Joon Weon; Choi, Yeon Seok

2013-01-01

In the present work, fast pyrolysis of Miscanthus sinensis was performed and the product yields and properties of the resulting biocrude oil were determined for varying reactor configurations and pyrolysis temperatures. Two types of reactors (rectangular and cylindrical fluidized beds) were adopted, and pyrolysis temperature was increased from 400 °C to 550 °C. Based on the results, it was found that the reaction temperature greatly influenced the product yield and the characteristics of biocrude oil. The highest yield of biocrude oil for the rectangular reactor was 48.9 wt.%, produced at 500 °C, and the highest yield for the cylindrical reactor was 50.01 wt.%, produced at 450 °C. Additionally, the biocrude oil yield in the rectangular reactor sharply decreased when reaction temperature was increased to 550 °C, while only a slight decrease was observed in the cylindrical reactor. From GC/MS analysis, biocrude oil was found to contain various chemical components, such as nonaromatic ketones, furans, sugars, lignin-derived phenols, guaiacols and syringols. In particular, the sugar content of the biocrude oil produced in rectangular reactor (2.11–9.35 wt.%) was generally lower than that produced in the cylindrical reactor (7.93–10.79 wt.%). - Highlights: • Fast pyrolysis of Miscanthus sinensis was performed in two fluidized bed reactors to obtain biocrude oil. • The yield and characteristics of the biocrude oil were scrutinized with changing reaction temperature and reactor type. • The reaction temperature was found to be the most influencing parameter for the fast pyrolysis reaction. • The different heating rate caused by reactor type has an effect on the final product yield and characteristics

Temperature stabilisation in Fischer–Tropsch reactors using phase change material (PCM)

International Nuclear Information System (INIS)

Odunsi, Ademola O.; O'Donovan, Tadhg S.; Reay, David A.

2016-01-01

The Fischer–Tropsch (FT) reaction is highly exothermic. The exothermicity combined with a high sensitivity of product selectivity to temperature constitute the main challenges in the design of FT reactors. Temperature control is particularly critical to the process in order to ensure longevity of the catalyst, optimise the product distribution, and to ensure thermo-mechanical reliability of the entire process. The use of encapsulated, Phase Change Material (PCM), in conjunction with a supervisory temperature control mechanism, could help mitigate these challenges and intensify the heat transport from the reactor. A 2D-axisymmetric, pseudo-homogeneous, steady-state model, with the dissipation of the enthalpy of reaction into an isothermal PCM sink, in a wall-cooled, single-tube fixed bed reactor is presented. Effective temperature control shows a shift in thermodynamic equilibrium, favouring the selectivity of longer chain hydrocarbons (C_5_+) to the disadvantage of CH_4 selectivity-a much desired outcome in the hydrocarbon Gas-to-Liquid (GTL) industry. - Highlights: • Phase change material is used to control temperature in a Fischer–Tropsch reactor. • Effective temperature control favours the production of C_5_+ over CH_4. • A 2D-axisymmetric, steady-state model is presented. • The model is verified against similar experimental work done in literature.

Flow instability tests for a particle bed reactor nuclear thermal rocket fuel element

Science.gov (United States)

Lawrence, Timothy J.

1993-05-01

Recent analyses have focused on the flow stability characteristics of a particle bed reactor (PBR). These laminar flow instabilities may exist in reactors with parallel paths and are caused by the heating of the gas at low Reynolds numbers. This phenomena can be described as follows: several parallel channels are connected at the plenum regions and are stabilized by some inlet temperature and pressure; a perturbation in one channel causes the temperature to rise and increases the gas viscosity and reduces the gas density; the pressure drop is fixed by the plenum regions, therefore, the mass flow rate in the channel would decrease; the decrease in flow reduces the ability to remove the energy added and the temperature increases; and finally, this process could continue until the fuel element fails. Several analyses based on different methods have derived similar curves to show that these instabilities may exist at low Reynolds numbers and high phi's ((Tfinal Tinitial)/Tinitial). These analyses need to be experimentally verified.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

Science.gov (United States)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Secret high-temperature reactor concept for inertial fusion

International Nuclear Information System (INIS)

Monsler, M.J.; Meier, W.R.

1983-01-01

The goal of our SCEPTRE project was to create an advanced second-generation inertial fusion reactor that offers the potential for either of the following: (1) generating electricity at 50% efficiency, (2) providing high temperature heat (850 0 C) for hydrogen production, or (3) producing fissile fuel for light-water reactors. We have found that these applications are conceptually feasible with a reactor that is intrinsically free of the hazards of catastrophic fire or tritium release

Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

International Nuclear Information System (INIS)

Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

2011-01-01

Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

Graphite for high-temperature reactors

International Nuclear Information System (INIS)

Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

1976-01-01

The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

High-temperature reactors: a recent past, a near future

International Nuclear Information System (INIS)

Ballot, B.

2007-01-01

While high-temperature reactors did experience major developments in the past, in Europe in particular, significant R and D efforts are required, if a major innovation deployment is to be made possible, of modular reactors having the capability of being coupled, in reliable, economic fashion, to an industrial process. The aim: the construction, before the next decade is out more swiftly than is feasible for other fourth-generation systems of an industrial prototype, coupled to such a process. The Areva Group takes up this approach, with its ANTARES project. For the purposes of characterizing the thermal properties of heterogeneous, multi-scale materials, as a function of temperature, experimental and numerical instruments have been developed at the Microstructure and Behavior Laboratory (Laboratoire microstructure et comportement), at CEA Le Ripault Center. They have been applied to the thermal characterization of the various layers in a high temperature reactor (HTR) fuel ball. (authors)

Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

International Nuclear Information System (INIS)

Stroh, K.R.

1979-03-01

The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

IAEA high temperature gas-cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2000-01-01

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

IAEA high temperature gas cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

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

[Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

Science.gov (United States)

Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

2010-03-01

Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor

International Nuclear Information System (INIS)

Du Toit, C.G.; Rousseau, P.G.; Greyvenstein, G.P.; Landman, W.A.

2006-01-01

The theoretical basis and conceptual formulation of a comprehensive reactor model to simulate the thermal-fluid phenomena of the PBMR reactor core and core structures is given. Through a rigorous analysis the fundamental equations are recast in a form that is suitable for incorporation in a systems CFD code. The formulation of the equations results in a collection of one-dimensional elements (models) that can be used to construct a comprehensive multi-dimensional network model of the reactor. The elements account for the pressure drop through the reactor; the convective heat transport by the gas; the convection heat transfer between the gas and the solids; the radiative, contact and convection heat transfer between the pebbles and the heat conduction in the pebbles. Results from the numerical model are compared with that of experiments conducted on the SANA facility covering a range of temperatures as well as two different fluids and different heating configurations. The good comparison obtained between the simulated and measured results show that the systems CFD approach sufficiently accounts for all of the important phenomena encountered in the quasi-steady natural convection driven flows that will prevail after critical events in a reactor. The fact that the computer simulation time for all of the simulations was less than three seconds on a standard notebook computer also indicates that the new model indeed achieves a fine balance between accuracy and simplicity. The new model can therefore be used with confidence and still allow quick integrated plant simulations. (authors)

10 years of current generation in the AVR reactor. The high-temperature pebble-bed reactor - a hot tip for our future

Energy Technology Data Exchange (ETDEWEB)

Schenk, P [Arbeitsgemeinschaft Versuchs-Reaktor G.m.b.H., Juelich (Germany, F.R.); A G, Stadtwerke Duesseldorf [Germany, F.R.; Vereinigung Deutscher Elektrizitaetswerke e.V. (VDEW), Frankfurt am Main (Germany, F.R.)); Nehrling, H [Ministerium fuer Wirtschaft, Mittelstand und Verkehr des Landes Nordrhein-Westfalen, Duesseldorf (Germany, F.R.); Daeunert, U [Bundesministerium fuer Forschung und Technologie, Bonn-Bad Godesberg (Germany, F.R.); Schulten, R [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung; Mattick, W [Brown, Boveri und Cie A.G., Mannheim (Germany, F.R.)

1978-02-01

On 17th December 1967, the experimental nuclear power plant (AVR) in Juelich supplied RWE network with power for the first time. With the start of the power operation of the first German high-temperature reactor (HTR), a milestone was reached in the development of this new and progressive line of construction. On the same day exactly 10 years later, the successful work with the hottest nuclear reactor in the world was reviewed in the presence of 15 associates of the Arbeitsgemeinschaft Versuchsreaktor (AVR) Ltd. and the personnel of the experimental nuclear power plant at a festival event in the main auditorium of the nuclear power plant at Juelich before some 300 guests from central and local government, the board of control, representatives of the population of the Dueren area and the town of Juelich, as well as bodies of the power producing industry.

Temperature and Doppler coefficients of various space nuclear reactors

International Nuclear Information System (INIS)

Mughabghab, S.F.; Ludewig, H. Schmidt, E.

1993-01-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system

Temperature and Doppler Coefficients of Various Space Nuclear Reactors

Science.gov (United States)

Mughabghab, Said F.; Ludewig, Hans; Schmidt, Eldon

1994-07-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system.

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Dong Yujie; Scherer, Winfried

2005-01-01

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

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

International Nuclear Information System (INIS)

1978-01-01

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

Core homogenization method for pebble bed reactors

International Nuclear Information System (INIS)

Kulik, V.; Sanchez, R.

2005-01-01

This work presents a core homogenization scheme for treating a stochastic pebble bed loading in pebble bed reactors. The reactor core is decomposed into macro-domains that contain several pebble types characterized by different degrees of burnup. A stochastic description is introduced to account for pebble-to-pebble and pebble-to-helium interactions within a macro-domain as well as for interactions between macro-domains. Performance of the proposed method is tested for the PROTEUS and ASTRA critical reactor facilities. Numerical simulations accomplished with the APOLLO2 transport lattice code show good agreement with the experimental data for the PROTEUS reactor facility and with the TRIPOLI4 Monte Carlo simulations for the ASTRA reactor configuration. The difference between the proposed method and the traditional volume-averaged homogenization technique is negligible while only one type of fuel pebbles present in the system, but it grows rapidly with the level of pebble heterogeneity. (authors)

Pebble Bed Reactor Dust Production Model

Energy Technology Data Exchange (ETDEWEB)

Abderrafi M. Ougouag; Joshua J. Cogliati

2008-09-01

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

Pebble Bed Reactor Dust Production Model

International Nuclear Information System (INIS)

Abderrafi M. Ougouag; Joshua J. Cogliati

2008-01-01

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production

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

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

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

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

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

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

Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

International Nuclear Information System (INIS)

Li Junfeng

2016-01-01

High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

Assessment of very high temperature reactors in process applications

International Nuclear Information System (INIS)

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

1976-01-01

In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

Trickle bed reactor model to simulate the performance of commercial diesel hydrotreating unit

Energy Technology Data Exchange (ETDEWEB)

C. Murali; R.K. Voolapalli; N. Ravichander; D.T. Gokak; N.V. Choudary [Bharat Petroleum Corporation Ltd., Udyog Kendra (India). Corporate R& amp; D Centre

2007-05-15

A two phase mathematical model was developed to simulate the performance of bench scale and commercial hydrotreating reactors. Major hydrotreating reactions, namely, hydrodesulphurization, hydrodearomatization and olefins saturation were modeled. Experiments were carried out in a fixed bed reactor to study the effect of different process variables and these results were used for estimating kinetic parameters. Significant amount of feed vaporization (20-50%) was estimated under normal operating conditions of DHDS suggesting the importance of considering feed vaporization in DHDS modeling. The model was validated with plant operating data, under close to ultra low sulphur levels by correctly accounting for feed vaporization in heat balance relations and appropriate use of hydrodynamic correlations. The model could predict the product quality, reactor bed temperature profiles and chemical hydrogen consumption in commercial plant adequately. 14 refs., 7 figs., 6 tabs.

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.

The United States fluoride-salt-cooled high-temperature reactor program

International Nuclear Information System (INIS)

Holcomb, David E.

2013-01-01

The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. 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 show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

Physics of high-temperature reactors

International Nuclear Information System (INIS)

Massimo, L.

1976-01-01

The subject is covered in chapters entitled: general description of the HTR core; general considerations about reactor physics; neutron cross-sections; basic aspects of transport and diffusion theory; methods for the solution of the diffusion equation; slowing-down and thermalization in graphite; resonance absorption; spectrum calculations and cross-section averaging; burn-up; core design; fuel management and cost calculations; temperature coefficient; core dynamics and accident analysis; reactor control; peculiarities of HTR physics; analysis of calculational accuracy; sequence of reactor design calculations. (U.K.)

Safety Philosophy in Process Heat Plants Coupled to High Temperature Reactors

International Nuclear Information System (INIS)

Brown, Nicholas R.; Revankar, Shripad T.

2011-01-01

With the future availability of fossil fuel resources in doubt, high temperature nuclear reactors have the potential to be an important technology in the near term. Due to a high coolant outlet temperature, high temperature reactors (HTR) can be used to drive chemical plants that directly utilize process heat. Additionally, the high temperature improves the thermodynamic efficiency of the energy utilization. Many applications of high temperature reactors exist as a thermal driving vector for endothermic chemical process plants. Hydrogen generation using the General Atomics (GA) sulfur iodine (SI) cycle is one promising application of high temperature nuclear heat. The main chemical reactions in the SI cycle are: 1. I 2 +SO 2 + 2H 2 O → 2HI + H 2 SO 4 (Bunsen reaction) 2. H 2 SO 4 → H 2 O + SO 2 + 1/2O 2 (Sulfuric acid decomposition) 3. 2HI → H 2 + I 2 (Hydrogen Iodide decomposition). With the exception of hydrogen and oxygen, all relevant reactants are recycled within the process. However, there are many unresolved safety and operational issues related to implementation of such a coupled plant

Medium-size high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

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

1980-08-01

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

Analysis of startup strategies for a particle bed reactor nuclear rocket engine

Science.gov (United States)

Suzuki, D. E.

1993-06-01

This paper develops and analyzes engine system startup strategies for a particle bed reactor (PBR) nuclear rocket engine. The strategies are designed to maintain stable flow through the PBR fuel element while reaching the design conditions as quickly as possible. The analyses are conducted using a computer model of a representative particle bed reactor and engine system. Elements of the startup strategy considered include: the coordinated control of reactor power and coolant flow; turbine inlet temperature and flow control; and use of an external starter system. The simulation results indicate that the use of an external starter system enables the engine to reach design conditions very quickly while maintaining the flow well away from the unstable regime. If a bootstrap start is used instead, the transient does not progress as fast and approaches closer to the unstable flow regime, but allows for greater engine reusability. These results can provide important information for engine designers and mission planners.

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.

Experimental studies on the coolability of packed beds. Flooding of hot dry packed beds

International Nuclear Information System (INIS)

Leininger, S.; Kulenovic, R.; Laurien, E.

2013-01-01

In case of a severe accident in a nuclear power plant meltdown of the reactor core can occur and form a packed bed in the lower plenum of the reactor pressure vessel (RPV) after solidification due to contact with water. The removal of after-heat and the long-term coolability is of essential interest. The efficient injection of cooling water into the packed bed has to be assured without endangering the structural integrity of the reactor pressure vessel. The experiments performed aimed to study the dry-out and the quenching (flooding) of hot dry packed beds. Two different inflow variants, bottom- and top-flooding including the variation of the starting temperature of the packed bed and the injection rate were studied. In case of bottom flooding the quenching time increases with increasing packed bed temperature and decreasing injection rate. In case of top flooding the flow pattern is more complex, in a first phase the water flows preferentially toward the RPV wall, the flow paths conduct the water downwards. The flow resistance of the packed bed increases with increasing bed temperatures. The quenching temperatures increase significantly above average.

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

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

1990-01-01

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

Effect of process temperature on morphology of CNTs grown in a vertically fluidized bed reactor with Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst

Energy Technology Data Exchange (ETDEWEB)

Shukrullah, Shazia, E-mail: zshukrullah@gmail.com; Mohamed, Norani Muti, E-mail: noranimuti-mohamed@petronas.com.my; Shaharun, Maizatul Shima, E-mail: maizats@petronas.com.my [Department of Fundamental and Applied Sciences, Universiti Teknologi PETRONAS, 31750 Tronoh, Perak (Malaysia)

2015-07-22

Carbon nanotubes (CNTs) are one of the most researched materials due to their exceptional mechanical and electrical properties. Among the various techniques, catalytic chemical vapor deposition in a fluidized bed reactor is the most promising technique for bulk production of CNTs. To meet the demand of good quality along with the bulk production of CNTs, the effect of reaction temperature on the micro structures, morphology, diameter, quality and quantity of CNTs was investigated in these studies. CNTs were synthesized at process temperature ranging from 700-850°C by catalytic decomposition of C{sub 2}H{sub 4} on Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst a vertical fluidized bed reactor. The microstructures of the grown CNTs at different reaction temperatures were investigated by using scanning electron microscope. The results of this study depicted a positive correlation between the average diameter of CNTs and reaction temperature. Narrow diameters (35∼40 nm) of CNTs with fewer defects were found at the low and mild temperatures, in particular 800°C. At this temperature, a dynamic equilibrium between the rate of C{sub 2}H{sub 4} decomposition and CNTs quantity was found due to maximum carbon diffusion over catalyst. The CNTs produced with Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst wer e also exhibiting high quality with relatively small mean outer diameter and fewer surface defects.

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

High-temperature-structural design and research and development for reactor system components

International Nuclear Information System (INIS)

Matsumura, Makoto; Hada, Mikio

1985-01-01

The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

Research and development for high temperature gas cooled reactor in Japan

International Nuclear Information System (INIS)

Taketani, K.

1978-01-01

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

Role and status of scaled experiments in the development of fluoride-salt-cooled, high-temperature reactors - 15185

International Nuclear Information System (INIS)

Zweibaum, N.; Huddar, L.; Laufer, M.R.; Peterson, P.F.; Hughes, J.T.; Blandford, E.D.; Scarlat, R.O.

2015-01-01

Development of fluoride-salt-cooled, high-temperature reactor (FHR) technology requires a better understanding of key hydrodynamic and heat transfer phenomena associated with this novel class of reactors. The use of simulant fluids that can match the most important non dimensional numbers between scaled experiments and prototypical FHR systems enables integral effects tests (IETs) to be performed at reduced cost and difficulty for FHR code validation. The University of California at Berkeley (UCB) and the University of New Mexico (UNM) have built a number of IETs and separate effects tests to investigate pebble-bed FHR (PB-FHR) phenomenology using water or simulant oils such as Dowtherm A. PB-FHR pebble motion and porous media flow dynamics have been investigated with UCB's pebble recirculation experiments using water and plastic spheres. Transient flow of high-Prandtl-number fluids around hot spheres has also been investigated by UCB to measure Nusselt numbers in pebble-bed cores, using simulant oils and copper spheres. Finally, single-phase forced/natural circulation has been investigated using the scaled height, reduced flow area loops of the Compact Integral Effects Test facility at UCB and a multi-flow regime loop at UNM, using Dowtherm A oil. The scaling methodology and status of these ongoing experiments are described here

Deposition reactors for solar grade silicon: A comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

Science.gov (United States)

Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.

2015-12-01

Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Process Stability Identification Through Dynamic Study of Single-bed Ammonia Reactor with Feed-Effluent Heat Exchanger (FEHE

Directory of Open Access Journals (Sweden)

Adhi Tri Partono

2018-01-01

Full Text Available In ammonia reactor system, a feed-effluent heat exchanger (FEHE is typically installed to utilize reaction-generated heat to heat the reactor’s feed. Utilizing energy from exothermic reaction to the incoming feed stream is often called “autothermic operation”. Despite the advantage of FEHE, there is a risk of utilizing FEHE in a reactor system such as instability of process temperature or known as hysteresis. Hysteresis phenomena in chemical process could cause operational problems, for example it could damage the integrity of the equipment’s material. This paper aims to evaluate the dynamic behavior of a single-bed ammonia reactor with FEHE, particularly to propose a way to prevent instability within the system. The dynamic simulation of the single-bed ammonia reactor with FEHE was performed with Aspen HYSYS v8.8. The result of the simulation result shows that hysteresis phenomenon in the ammonia reactor system occurs when the feed’s temperature is below a certain value. If the feed temperature reaches that value, the temperature of the reactor’s outlet oscillates. One of the solution to keep the feed temperature above that critical value is by installing a trim heater within the system. Based on the simulation, trim heater installation within the system is able to prevent hysteresis in the system evaluated.

The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

International Nuclear Information System (INIS)

Forsberg, C.W.; Peterson, P.F.; Ott, L.

2004-01-01

Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

Materials for high temperature reactor vessels

International Nuclear Information System (INIS)

Buenaventura Pouyfaucon, A.

2004-01-01

Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

A Numerical Model for Trickle Bed Reactors

Science.gov (United States)

Propp, Richard M.; Colella, Phillip; Crutchfield, William Y.; Day, Marcus S.

2000-12-01

Trickle bed reactors are governed by equations of flow in porous media such as Darcy's law and the conservation of mass. Our numerical method for solving these equations is based on a total-velocity splitting, sequential formulation which leads to an implicit pressure equation and a semi-implicit mass conservation equation. We use high-resolution finite-difference methods to discretize these equations. Our solution scheme extends previous work in modeling porous media flows in two ways. First, we incorporate physical effects due to capillary pressure, a nonlinear inlet boundary condition, spatial porosity variations, and inertial effects on phase mobilities. In particular, capillary forces introduce a parabolic component into the recast evolution equation, and the inertial effects give rise to hyperbolic nonconvexity. Second, we introduce a modification of the slope-limiting algorithm to prevent our numerical method from producing spurious shocks. We present a numerical algorithm for accommodating these difficulties, show the algorithm is second-order accurate, and demonstrate its performance on a number of simplified problems relevant to trickle bed reactor modeling.

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

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

International Nuclear Information System (INIS)

Ingersoll, D.T.

2004-01-01

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

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

International Nuclear Information System (INIS)

1978-09-01

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

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

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.

1976-12-01

In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

Experimental investigation of pyrolysis of rice straw using bench-scale auger, batch and fluidized bed reactors

International Nuclear Information System (INIS)

Nam, Hyungseok; Capareda, Sergio C.; Ashwath, Nanjappa; Kongkasawan, Jinjuta

2015-01-01

Energy conversion efficiencies of three pyrolysis reactors (bench-scale auger, batch, and fluidized bed) were investigated using rice straw as the feedstock at a temperature of 500 °C. The highest bio-oil yield of 43% was obtained from the fluidized bed reactor, while the maximum bio-char yield of 48% was obtained from the batch reactor. Similar bio-oil yields were obtained from the auger and batch type reactors. The GCMS and FTIR were used to evaluate the liquid products from all reactors. The best quality bio-oil and bio-char from the batch reactor was determined to have a heating value of 31 MJ/kg and 19 MJ/kg, respectively. The highest alkali mineral was found in the bio-char produced from the auger reactor. The energy conversion efficiencies of the three reactors indicated that the majority of the energy (50–64%) was in the bio-char products from the auger and batch reactors, while the bio-oil from the fluidized bed reactor contained the highest energy (47%). A Sankey diagram has been produced to show the flows of product energy from each pyrolysis process. The result will help determine which conversion process would be optimal for producing specific products of bio-char, bio-oil, and gas depending on the needs. - Highlights: • Pyrolysis products from auger, batch, and fluidized bed reactor were examined. • O/C ratios of bio-oils stayed in specific ranges depending on the process reactors. • The largest quantity of bio-oil from fluidized, while the best quality from batch. • The highest alkali concentration of 37 g/kg included in the auger based bio-char. • Sankey diagram was used to understand the energy distribution from reactors.

Assessment of very high-temperature reactors in process applications

International Nuclear Information System (INIS)

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 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 0 F is attainable with near-term technology. However, process heat in excess of 1600 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

Radioisotope tracer study in a pilot-scale trickle bed reactor

Energy Technology Data Exchange (ETDEWEB)

Pant, H.J.; Pendharkar, A.S. [Bhabha Atomic Research Centre, Isotope Applications Div, Bombay (India); Prasad, J.S.; Maiti, R.N.; Chawla, R.; Lahri, R.N.; Ram Babu, D. [Engineers India Limited, Gurgaon, Haryana (India); Berne, Ph. [CEA Grenoble, DTEN/SAT, 38 (France)

2001-07-01

Trickle bed reactor (TBR) is a reactor in which a liquid and a gas phase flow concurrently downwards through a fixed bed of catalyst particles while the reaction takes place. The trickle bed reactors find a number of applications in petroleum refining, chemical. petro-chemical and bio-chemical processes due to their suitability for hydro-processing operations.The knowledge of radial distribution, mean residence time (MRT). liquid holdup and degree of axial mixing is a basic requirement to evaluate the reactor performance of a TBR, its optimal size, the physical and chemical interactions and the pumping requirements. Measurement and analysis of residence time distribution (RTD) of liquid phase facilitate the determination of these parameters. This paper describes the measurement of RTD of liquid phase in a pilot-scale trickle bed reactor operating at high pressure. Kerosene and nitrogen were used as liquid and gas phase, respectively. Porous alumina catalyst particles were used as packing. Bromine-82 as para-di-bromo benzene was used as a tracer to measure RTD of organic liquid phase. The RTD of kerosene was measured as a function of liquid and gas flow rates and pressure. The axial dispersion model with exchange between active and dead zones was proposed and used to simulate the measured RTD data and model parameters i.e mean residence time ({tau}), Peclet Number (Pe), dynamic fraction of the liquid ({phi}) and the number of transfer units (N) were obtained. The results of the simulations indicated intermediate degree of axial mixing of liquid phase. The RTD predicted by the axial dispersion model with exchange between active and dead zones fits well to the experimentally measured RTD data and thus is a suitable model to describe the dynamics of the liquid phase in TBRs filled with porous catalyst particles. (authors)

Hydrolysis of cellulose in a cellulase-bead fluidized bed reactor

Energy Technology Data Exchange (ETDEWEB)

Karube, I; Tanaka, S; Shirai, T; Suzuki, S

1977-08-01

Cellulase was immobilized in a collagen fibril matrix, and no leakage of cellulase from the collagen fibril matrix was observed. The immobilized cellulase was more stable than the native cellulase. The substrate cellulose was hydrolyzed quantitatively with immobilized cellulase. The final reaction product was identified as glucose. Immobilized cellulase was used in a fluidized bed reactor where the pressure drop of the fluidized bed reactor was low and constant. Cellulose was hydrolyzed to glucose by the cellulase-bead fluidized bed reactor. The minimum flow velocity (U/sub mf/) was 0.5 cm/sec and the optimum flow velocity of the cellulose hydrolysis was 1 cm/sec.

Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

2009-01-01

The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

Review of PSI studies on reactor physics and thermal fluid dynamics of pebble bed reactors

International Nuclear Information System (INIS)

Prasser, Horst-Michael

2014-01-01

Switzerland is member of the Generation IV International Forum (GIF). The related work takes entirely place at PSI in the working groups of Gas-Cooled Fast Reactors and Very High Temperature Reactors. In the past, PSI has performed experimental and theoretical studies on criticality issues of pebble beds at the PROTEUS reactor, as well as a preliminary risk assessment of a prototypal HTR as an input for a comparison of energy supply options. PROTEUS was a critical assembly with an annular driver zone. The central region was filled by arrangements of fuel spheres. The reactivity effect of a water ingress was investigated by simulating the water by polyethylene rods of different diameter inserted into the gaps of a regular package. For sub-criticality measurements in pebble beds, a built-in pulsed neutron source was used. The experimental results were used to validate diffusion and higher order neutron transport models. Concerning thermal hydraulics of gas flows, the vast experience of PSI is focused on hydrogen transport, accumulation, and dispersion in containments of light water reactors. The phenomena are comparable in many aspects to the fluid dynamic issues relevant to HTR. Experiments on hydrogen flows are performed for numerous scenarios in the large-scale containment test facility PANDA. Hydrogen is substituted by helium as a model fluid. An important generic aspect is turbulent mixing in the presence of strong stratification, which is relevant for HTR as well. In a parallel project, generic small-scale mixing experiments with a high density ratio of 1:7 are carried out in a horizontal rectangular channel, where helium and nitrogen flows are brought into contact downstream of the rear edge of a splitter plate. Due to the high density ratio, turbulent mixing is affected by strong non-Boussinesq effects. The measurements taken by Particle Imaging Velocimetry (PIV) and Laser Induced Fluorescence techniques are compared to RANS and LES simulations. Similar large

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

Energy Technology Data Exchange (ETDEWEB)

Woods, Brian; Gutowska, Izabela; Chiger, Howard

2018-03-26

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion to ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with

Rotating bed reactor for CLC: Bed characteristics dependencies on internal gas mixing

International Nuclear Information System (INIS)

Håkonsen, Silje Fosse; Grande, Carlos A.; Blom, Richard

2014-01-01

Highlights: • A mathematical model for the rotating CLC reactor has been developed. • The model reflects the gas distribution in the reactor during CLC operation. • Radial dispersion in the rotating bed is the main cause for internal gas mixing. • The model can be used to optimize the reactor design and particle characteristics. - Abstract: A newly designed continuous lab-scale rotating bed reactor for chemical looping combustion using CuO/Al 2 O 3 oxygen carrier spheres and methane as fuel gives around 90% CH 4 conversion and >90% CO 2 capture efficiency based on converted methane at 800 °C. However, from a series of experiments using a broad range of operating conditions potential CO 2 purities only in the range 20–65% were yielded, mostly due to nitrogen slip from the air side of the reactor into the effluent CO 2 stream. A mathematical model was developed intending to understand the air-mixing phenomena. The model clearly reflects the gas slippage tendencies observed when varying the process conditions such as rotation frequency, gas flow and the flow if inert gas in the two sectors dividing the air and fuel side of the reactor. Based on the results, it is believed that significant improvements can be made to reduce gas mixing in future modified and scaled-up reactor versions

Fuel arrangement for high temperature gas cooled reactor

International Nuclear Information System (INIS)

Tobin, J.M.

1978-01-01

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

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.

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

International Nuclear Information System (INIS)

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 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 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

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-01-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

An energy amplifier fluidized bed nuclear reactor concept

International Nuclear Information System (INIS)

Sefidvash, F.; Seifritz, W.

2001-01-01

The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle. (orig.) [de

Operational methods of the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Borges, V.; Sefidvash, F.

1993-01-01

The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

International Nuclear Information System (INIS)

Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

2015-01-01

The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ −1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ −1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ −1 higher, respectively, than in the PBD reactor at 53 J l −1 . The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment. (paper)

Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

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Subagjo

2014-07-01

Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

International Nuclear Information System (INIS)

Elliott, H.H.; Holloway, J.H.; Abbott, D.G.

1979-01-01

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