Sample records for burner reactor systems

  1. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire


    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  2. Advanced burner test reactor preconceptual design report.

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    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA


    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  3. Advanced Burner Reactor Preliminary NEPA Data Study.

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    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division


    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  4. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.


    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  5. Advanced Burner Reactor 1000MWth Reference Concept

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kellogg, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Salev, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Seidensticker, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Tang, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Tzanos, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Chikazawa, Y. [Argonne National Lab. (ANL), Argonne, IL (United States)


    The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence, to validate the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat.

  6. Preliminary safety evaluation of the advanced burner test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division


    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  7. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division


    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition


    Sonotech, Inc. (Sonotech), of Atlanta, GA, the developer of the Cello® pulse combustion burner, claims that its burner system can be beneficial to a variety of combustion processes. The system incorporates a combustor that can be tuned to induce large amplitude sonic pulsation...

  9. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein


    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  10. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis


    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  11. Use of freeze-casting in advanced burner reactor fuel design

    Energy Technology Data Exchange (ETDEWEB)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)


    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

  12. Rapid starting methanol reactor system (United States)

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.


    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  13. Development of KB-200. Kawasaki digital burner control system

    Energy Technology Data Exchange (ETDEWEB)

    Miyayama, Toshio; Hoshino, Norihiko; Nakamori, Hideki; Wasada, Norihiko


    Control equipments of a thermal generation plant are making progress in getting higher in level, complex and integrated digitalization. Concerning the digital control technology, as micro-electronics technology progresses, programmable controller which excels in flexibility and expandability of software is now widely popularized. Recently, however, higher level of reliability, maintenance ability and safety are required as in the discovery of abnormality and fact-finding or the protection of human error. To respond to these, Kawasaki Heavy Industries, in cooperationwith Fuji Electric Co., has developed a distributed digital burner control unit on the basis of the recent programmable controller. This report outlines the system control functions, construction, measures for high reliability, controllers, and system maintenance, etc.. (8 figs, 2 tabs, 2 refs)

  14. Multifuel burners based on the porous burner technology for the application in fuel cell systems; Mehrstofffaehige Brenner auf Basis der Porenbrennertechnik fuer den Einsatz in Brennstoffzellensystemen

    Energy Technology Data Exchange (ETDEWEB)

    Diezinger, S.


    The present doctoral thesis describes the development of multifuel burners based on the porous burner technology for the application in hydrocarbon driven fuel cell systems. One objective of such burners is the heating of the fuel cell system to the operating temperature at the cold start. In stationary operation the burner has to postcombust the waste gases from the fuel cell and the gas processing system in order to reduce the pollutant emissions. As the produced heat is required for endothermal processes like the steam reforming the burner has a significant influence on the system's efficiency. The performed investigations are targeting on a gasoline driven PEMFC-System with steam reforming. In such systems the burner has to be capable to combust the system's fuel gasoline at the cold start, a low calorific fuel cell offgas (HU = 6,4 MJ/kg) in stationary operation and a hydrogen rich gas in the case of an emergency shut down. Pre-tests revealed that in state of the art porous burners the flame front of hydrogen/air combustion can only be stabilized at very high excess air ratios. In basic investigations concerning the stabilization of flame fronts in porous media the dominant influence parameters were determined. Based on this findings a new flame trap was developed which increases the operational range with hydrogen rich mixtures significantly. Furthermore the burning velocity at stationary combustion in porous media was investigated. The dependency of the porous burning velocity on the excess air ratio for different hydrocarbons and hydrogen as well as for mixtures of both was determined. The results of these basic investigations were applied for the design of a multifuel burner. In order to achieve an evaporation of the gasoline without the use of additional energy, an internal heat exchanger section for heating the combustion air was integrated into the burner. Additionally different experimental and numerical methods were applied for designing the

  15. Actinide transmutation using inert matrix fuels versus recycle in a low conversion fast burner reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, M.R.; Schneider, E.A.; Recktenwald, G.; Cady, K.B. [The Department of Mechanical Engineering, The University of Texas at Austin, 1 University Station, C2200, Austin, 78712 (United States)


    would require an infinite fuel residence time. In previous work we have shown that the amount of fluence required to achieve a unit of burnup in yttrium stabilized ZrO{sub 2} based IMF with 85 w/o zirconium oxide and 15 w/o minor actinides (MA) and plutonium increases dramatically beyond 750 MWd/kgIHM (75% burnup). In this paper we discuss the repository implications for recycle of actinides in LWR's using this type of IMF and compare this to actinide recycle in a low conversion fast burner reactor. We perform the analysis over a finite horizon of 100 years, in which reprocessing of spent LWR fuel begins in 2020. Reference [1] C. Lombardi and A. Mazzola, Exploiting the plutonium stockpiles in PWRs by using inert matrix fuel, Annals of Nuclear Energy. 23 (1996) 1117-1126. [2] U. Kasemeyer, J.M. Paratte, P. Grimm and R. Chawla, Comparison of pressurized water reactor core characteristics for 100% plutonium-containing loadings, Nuclear Technology. 122 (1998) 52-63. [3] G. Ledergerber, C. Degueldre, P. Heimgartner, M.A. Pouchon and U. Kasemeyer, Inert matrix fuel for the utilisation of plutonium, Progress in Nuclear Energy. 38 (2001) 301-308. [4] U. Kasemeyer, C. Hellwig, J. Lebenhaft and R. Chawla, Comparison of various partial light water reactor core loadings with inert matrix and mixed oxide fuel, Journal of Nuclear Materials. 319 (2003) 142-153. [5] E.A. Schneider, M.R. Deinert and K.B. Cady, Burnup simulations of an inert matrix fuel using a two region, multi-group reactor physics model, in Proceedings of the physics of advanced fuel cycles, PHYSOR 2006, Vancouver, BC, 2006. [6] E.A. Schneider, M.R. Deinert and K.B. Cady, Burnup simulations and spent fuel characteristics of ZRO{sub 2} based inert matrix fuels, Journal of Nuclear Materials. 361 (2007) 41-51. (authors)

  16. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)


    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  17. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Wade, D.C. [Argonne National Lab., IL (United States); Palmiotti, G. [CEA - Cadarache, Saint-Paul-Les-Durance (France)


    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted.

  18. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)


    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and


    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.


    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.


    Treshow, M.


    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  1. Evaluating the efficacy of a minor actinide burner

    Energy Technology Data Exchange (ETDEWEB)

    Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Omberg, R.P.; Wootan, D.W.


    The efficacy of a minor actinide burner can be evaluated by comparing safety and economic parameters to the support ratio. Minor actinide mass produced per unit time in this number of Light Water Reactors (LWRs) can be burned during the same time period in one burner system. The larger the support ratio for a given set of safety and economic parameters, the better. To illustrate this concept, the support ratio for selected Liquid Metal Reactor (LMR) burner core designs was compared with corresponding coolant void worths, a fundamental safety concern following the Chernobyl accident. Results can be used to evaluate the cost in reduced burning of minor actinides caused by LMR sodium void reduction efforts or to compare with other minor actinide burner systems.

  2. Numerical modelling of the CHEMREC black liquor gasification process. Conceptual design study of the burner in a pilot gasification reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marklund, Magnus


    The work presented in this report is done in order to develop a simplified CFD model for Chemrec's pressurised black liquor gasification process. This process is presently under development and will have a number of advantages compared to conventional processes for black liquor recovery. The main goal with this work has been to get qualitative information on influence of burner design for the gas flow in the gasification reactor. Gasification of black liquor is a very complex process. The liquor is composed of a number of different substances and the composition may vary considerably between liquors originating from different mills and even for black liquor from a single process. When a black liquor droplet is gasified it loses its organic material to produce combustible gases by three stages of conversion: Drying, pyrolysis and char gasification. In the end of the conversion only an inorganic smelt remains (ideally). The aim is to get this smelt to form a protective layer, against corrosion and heat, on the reactor walls. Due to the complexity of gasification of black liquor some simplifications had to be made in order to develop a CFD model for the preliminary design of the gasification reactor. Instead of modelling droplets in detail, generating gas by gasification, sources were placed in a prescribed volume where gasification (mainly drying and pyrolysis) of the black liquor droplets was assumed to occur. Source terms for the energy and momentum equations, consistent with the mass source distribution, were derived from the corresponding control volume equations by assuming a symmetric outflow of gas from the droplets and a uniform degree of conversion of reactive components in the droplets. A particle transport model was also used in order to study trajectories from droplets entering the reactor. The resulting model has been implemented in a commercial finite volume code (AEA-CFX) through customised Fortran subroutines. The advantages with this simple

  3. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L


    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  4. Industrial thermal oxidation with an innovative burner management system; Industrielle thermische Nachverbrennung mit innovativem Brenner-Managementsystem

    Energy Technology Data Exchange (ETDEWEB)

    Gnoss, T. [Siemens Building Technologies HVAC Product GmbH, Rastatt (Germany); Pilz, R. [Control and Heating-Systems, Felsberg-Gensungen (Switzerland); Saenger, P. [Siemens Building Technologies HVAC Product GmbH, Frankfurt am Main(Germany)


    In view of rising energy costs and the emission limits stipulated by the latest 'TA-Luft' (Technical Directive: Prevention of Air Pollution) and 'BImSchV' (Federal Immission Control Ordinance in force in Germany), industrial thermal oxidation plants must be either completely replaced or a new burner system must be installed to ensure compliance with the latest environmental standards that demand restriction of pollutant emissions. Replacement of the original burner control system by a state-of-the-art burner management system improves not only the combustion process and the flue gas quality but also saves energy and thus costs through the use of a thermal incinerator. One of the key features of a thermal oxidation plant is a new technology used for controlling and monitoring the burner. The following article examines the innovative LMV5.. burner management system which offers a host of functions, such as burner control, electronic fuel / air ratio control, valve proving and load control - components which, previously, had to be separately assembled and electrically interconnected. (orig.)

  5. Combustion systems: a porous-matrix burner and a surface combustor. Topical report, June 1984-December 1985

    Energy Technology Data Exchange (ETDEWEB)

    Jasionowski, W.J.; Kunc, W.; Khinkis, M.J.; Zawacki, T.S.


    In this study, two combustion systems were experimentally evaluated for potential application to gas-fired appliances: an atmospheric porous-matrix burner and a packed-bed surface combustor. The normal radiant output of a porous matrix burner was measured over a range of input from 30,000 to 110,000 Btu/h-sq ft and aerations from 0% to 40% excess air. The data were compared to similar published data for radiant tile burners. Emissions of nitric oxide (NO), nitrogen dioxide (NO/sub 2/), and carbon monoxide (CO) were also measured and reported. The potential advantage of the type of burner is its ability to produce high levels of radiation heat at high gas-loading levels (Btu/h-sq ft.).

  6. Burner (Stinger) (United States)

    ... feel overly tired. These are symptoms of a concussion . The doctor will ask you questions about what ... a Burner? Injuries to the brachial plexus can happen when a person's head is pushed forcefully down ...

  7. Nuclear reactor sealing system (United States)

    McEdwards, James A.


    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  8. Attrition reactor system (United States)

    Scott, Charles D.; Davison, Brian H.


    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  9. Multifuel burner

    Energy Technology Data Exchange (ETDEWEB)

    Raybould, J.D.


    A design is proposed for turbulent burner (B) for simultaneous burning of powder, liquid and/or gaseous fuel (F). The liquid F is sprayed with the help of a rotation sprayer arranged on the axis of the burner device. The gas can be supplied through the opening made in the dish-shaped bottom encompassing the central part of the B. The powder F (aeromixture) enters the combustion zone through the channels with vortex blades arranged on the periphery of the bottom of the B. Through the annular channel arranged around the rotation sprayer, primary air is supplied, and through the channels arranged on the periphery of the B, secondary air. The percentage of solid F during operation of the B can be 75-90%.

  10. Optimization of Fast Critical Experiments to Reduce Nuclear Data Uncertainties in Support of a Fast Burner Reactor Design Concept (United States)

    Stover, Tracy E., Jr.

    An optimization technique has been developed to select optimized experimental design specifications to produce data specifically designed to be assimilated to optimize a given reactor concept. Data from the optimized experiment is assimilated to generate posteriori uncertainties on the reactor concept's core attributes from which the design responses are computed. The reactor concept is then optimized with the new data to realize cost savings by reducing margin. The optimization problem iterates until an optimal experiment is found to maximize the savings. A new generation of innovative nuclear reactor designs, in particular fast neutron spectrum recycle reactors, are being considered for the application of closing the nuclear fuel cycle in the future. Safe and economical design of these reactors will require uncertainty reduction in basic nuclear data which are input to the reactor design. These data uncertainty propagate to design responses which in turn require the reactor designer to incorporate additional safety margin into the design, which often increases the cost of the reactor. Therefore basic nuclear data needs to be improved and this is accomplished through experimentation. Considering the high cost of nuclear experiments, it is desired to have an optimized experiment which will provide the data needed for uncertainty reduction such that a reactor design concept can meet its target accuracies or to allow savings to be realized by reducing the margin required due to uncertainty propagated from basic nuclear data. However, this optimization is coupled to the reactor design itself because with improved data the reactor concept can be re-optimized itself. It is thus desired to find the experiment that gives the best optimized reactor design. Methods are first established to model both the reactor concept and the experiment and to efficiently propagate the basic nuclear data uncertainty through these models to outputs. The representativity of the experiment

  11. Development of stoker-burner wood chip combustion systems for the UK market

    Energy Technology Data Exchange (ETDEWEB)



    The document makes a case for the development of a design of wood chip stoker-burner more suited to the UK than those currently imported from Sweden and Finland. The differences would centre on market conditions, performance and cost-effectiveness and the devices would be manufactured or part-manufactured in the UK. Econergy Limited was contracted by the DTI as part of its Sustainable Energy Programmes to design and construct an operational prototype stoker-burner rated at 120 kWth. A test rig was built to: (i) study modified burner heads and (ii) develop control hardware and a control strategy. Both (i) and (ii) are described. Tests brought about an increase in performance of the burner head and its wet wood performance. It was considered that further improvements are achievable and six areas for future study were suggested.

  12. Solvent refined coal reactor quench system (United States)

    Thorogood, R.M.


    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.


    Sonotech, Inc. (Sonotech) of Atlanta, Georgia, has developed a pulse combustion burner technology that claims to offer benefits when applied in a variety of combustion processes. The technology incorporates a combustor that can be tuned to induce large-amplitude acoustic or soni...

  14. Investigations of coal ignition in a short-range flame burner using optical measuring systems; Untersuchungen zur Kohlezuendung am Flachflammenbrenner unter Verwendung optischer Messtechnik

    Energy Technology Data Exchange (ETDEWEB)

    Hackert, G.; Kremer, H.; Wirtz, S. [Bochum Univ. (Germany). Lehrstuhl fuer Energieanlagentechnik


    The short-range flame burner and the KOALA reactor of DMT are experimental facilities for realistic simulation of coal conversion processes at high temperatures and pressures in atmospheric conditions. The TOSCA system enable measurements of temperatures, sizes, shapes and velocities of the fuel particles, which serve as a basis for a three-dimensional simulation model of coal combustion. In the future, further parameter studies will deepen the present knowledge of coal dust combustion under pressure and enable optimisation of the numerical models for simulation of industrial-scale systems for coal dust combustion under pressure. [Deutsch] Mit dem Flachflammenbrenner und dem KOALA-Reaktor der DMT stehen Versuchsapparaturen zur Verfuegung, mit deren Hilfe die Kohleumwandlungsprozesse bei hohen Temperaturen unter Druck und unter atmosphaerischen Bedingungen realistisch wiedergegeben werden. Das TOSCA-System erlaubt dabei die Bestimmung von Temperaturen, Groessen, Formen und Geschwindigkeiten der Brennstoffpartikel. Diese Daten liefern die Grundlage fuer die Erstellung eines dreidimensionalen Simulationsmodells zur Modellierung der Kohleverbrennung. In Zukunft werden weitere Parameterstudien das Verstaendnis der Kohlenstaubdruckverbrennung vertiefen und ein Optimierung der numerischen Modelle ermoeglichen, so dass die Simulation grosstechnischer Kohlenstaubdruckverbrennungsanlagen realisiert werden kann. (orig.)

  15. Automatically scramming nuclear reactor system (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.


    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  16. Ecothal burner development; Ecothal braennarutveckling

    Energy Technology Data Exchange (ETDEWEB)

    Lewin, Thomas [KANTHAL AB, Hallstahammar (Sweden)


    A SER burner system with catalytic cleaning have been optimised for an outer tube OD 100-115 mm. The aim has been to develop a burner with an emission of nitrogen oxides below 50 ppm and an efficiency higher than 80%. An optimised burner system have been realised but will not be stable enough for commercialisation. In order to fullfill the requirements it have to be regulated with closed loop oxygen sensor system regulating the air/gas supply (Lambda-value). Practically it is possible to reach 200-300 ppm nitrogen oxide with an efficiency around 70-80%. Following work have to focus on how to improve the stability considering geometrical changes when in operation but also towards accomodation of production tolerances and fluctuations in gas supply systems.

  17. Liquid metal cooled nuclear reactor plant system (United States)

    Hunsbedt, Anstein; Boardman, Charles E.


    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  18. Furnaces with multiple ?ameless combustion burners

    NARCIS (Netherlands)

    Danon, B.


    In this thesis three different combustion systems, equipped with either a single or multiple ?ameless combustion burner(s), are discussed. All these setups were investigated both experimentally and numerically, i.e., using Computational Fluid Dynamics (CFD) simulations. Flameless combustion is a

  19. Blending of hydrogen in natural gas distribution systems. Volume II. Combustion tests of blends in burners and appliances. Final report, June 1, 1976--August 30, 1977. [8, 11, 14, 20, 22, 25, and 31% hydrogen

    Energy Technology Data Exchange (ETDEWEB)



    The emerging ''hydrogen economy'' is a strong contender as one method to supplement or extend the domestic natural gas supply. This volume of the subject study ''Blending Hydrogen in Natural Gas Distribution Systems'' describes combustion studies to determine the maximum amount of hydrogen that can be blended in natural gas and utilized satisfactorily in typical appliances with no adjustment or conversion. Eleven pilot burners and twenty-three main burners typical of those in current use were operated on hydrogen-natural gas mixtures containing approximately 8, 11, 14, 20, 22, 25, and 31 percent, by volume, hydrogen. The eleven pilot burners and thirteen main burners were tested outside the appliance they were a part of. Ten main burners were tested in their respective appliances. Performance of the various burners tested are as follows: (1) Gas blends containing more than 6 to 11% hydrogen are the limiting mixtures for target type pilot burners. (2) Gas blends containing more than 20 to 22% hyrogen are the limiting mixtures for main burners operating in the open. (3) Gas blends containing more than 22 to 25% hydrogen are the limiting mixtures for main burners tested in appliances. (4) Modification of the orifice in target pilots or increasing the supply pressure to a minimum of 7 inches water column will permit the use of gas blends with 20% hydrogen.

  20. Fuel-flexible burner apparatus and method for fired heaters

    Energy Technology Data Exchange (ETDEWEB)

    Zink, Darton J.; Isaacs, Rex K.; Jamaluddin, A. S. (Jamal); Benson, Charles E.; Pellizzari, Roberto O.; Little, Cody L.; Marty, Seth A.; Imel, K. Parker; Barnes, Jonathon E.; Parker, Chris S.


    A burner apparatus for a fired heating system and a method of burner operation. The burner provides stable operation when burning gas fuels having heating values ranging from low to high and accommodates sudden wide changes in the Wobbe value of the fuel delivered to the burner. The burner apparatus includes a plurality of exterior fuel ejectors and has an exterior notch which extends around the burner wall for receiving and combusting a portion of the gas fuel. At least a portion of the hot combustion product gas produced in the exterior notch is delivered through channels formed in the burner wall to the combustion area at the forward end of the burner. As the Wobbe value of the gas fuel decreases, one or more outer series of addition ejectors can be automatically activated as needed to maintain the amount of heat output desired.

  1. Shutdown system for a nuclear reactor (United States)

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.


    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  2. Regenerative burner systems for batch furnaces in the steel industry; Regenerativbrenner fuer Doppel-P-Strahlheizrohre in einer Feuerverzinkungslinie

    Energy Technology Data Exchange (ETDEWEB)

    Georgiew, Alexander [Salzgitter Flachstahl GmbH, Salzgitter (Germany); Wuenning, Joachim G.; Bonnet, Uwe [WS Waermeprozesstechnik GmbH, Renningen (Germany)


    This article will describe the application of a new self regenerative burner in a continuous galvanizing line. After a brief introduction of the process line, the self regenerative burner will be described. Very high air preheat temperatures enable considerable energy savings and flameless oxidation suppresses the formation of NO{sub X}. (orig.)

  3. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.


    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  4. Evaluation of Gas Reburning & Low NOx Burners on a Wall Fired Boiler Performance and Economics Report Gas Reburning-Low NOx Burner System Cherokee Station Unit 3 Public Service Company of Colorado

    Energy Technology Data Exchange (ETDEWEB)

    None, None


    Under the U.S. Department of Energy's Clean Coal Technology Program (Round 3), a project was completed to demonstrate control of boiler NOX emissions and to a lesser degree, due to coal replacement, SO2 emissions. The project involved combining Gas Reburning with Low NOX Burners (GR-LNB) on a coal-fired electric utility boiler to determine if high levels of NOX reduction (70%) could be achieved. Sponsors of the project included the U.S. Department of Energy, the Gas Research Institute, Public Service Company of Colorado, Colorado Interstate Gas, Electric Power Research Institute, and the Energy and Environmental Research Corporation. The GR-LNB demonstration was performed on Public Service Company of Colorado's (PSCO) Cherokee Unit #3, located in Denver, Colorado. This unit is a 172 MW~ wall-fired boiler that uses Colorado Bituminous, low-sulfur coal. It had a baseline NOX emission level of 0.73 lb/106 Btu using conventional burners. Low NOX burners are designed to yield lower NOX emissions than conventional burners. However, the NOX control achieved with this technique is limited to 30-50%. Also, with LNBs, CO emissions can increase to above acceptable standards. Gas Reburning (GR) is designed to reduce NOX in the flue gas by staged fuel combustion. This technology involves the introduction of natural gas into the hot furnace flue gas stream. When combined, GR and LNBs minimize NOX emissions and maintain acceptable levels of CO emissions. A comprehensive test program was completed, operating over a wide range of boiler conditions. Over 4,000 hours of operation were achieved, providing substantial data. Measurements were taken to quantify reductions in NOX emissions, the impact on boiler equipment and operability and factors influencing costs. The GR-LNB technology achieved good NOX emission reductions and the goals of the project were achieved. Although the performance of the low NOX burners (supplied by others) was less than expected, a NOX reduction of

  5. Nuclear electric propulsion reactor control systems status (United States)

    Ferg, D. A.


    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  6. Fission control system for nuclear reactor (United States)

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  7. Cooling system for a nuclear reactor (United States)

    Amtmann, Hans H.


    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  8. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth


    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  9. Scanning tunneling microscope assembly, reactor, and system (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A


    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  10. Feasibility study of utilizing jatropha curcas oil as bio-diesel in an oil firing burner system (United States)

    Shaiful, A. I. M.; Jaafar, M. N. Mohd; Sahar, A. M.


    Jatropha oil derived from the Jatropha Curcas Linnaeus is one of the high potential plants to be use as bio-diesel. The purpose of this research is to carry out a feasibility study of using jatropha oil as bio-diesel on oil firing burner system. Like other bio-diesels, jatropha oil can also be used in any combustion engine and the performance and emissions such as NOx, SO2, CO and CO2 as well as unburned hydocarbon (UHC) from the engine will vary depending on the bio-diesel blends. The properties of Conventional Diesel Fuel (CDF) obtained will be used as baseline and the jatropha oil properties will be compared as well as other bio-diesels. From several researches, the properties of jatropha oil was found to be quite similar with other bio-diesel such as palm oil, neem, keranja and pongamia bio-diesel and complying with the ASTM standard for bio-diesel. Still, there are factors and issues concerning the use of jatropha oil such as technology, economy, legislation and resource. Plus, there several challenges to the growth of bio-diesel industry development since the world right now do not totally depend on the bio-diesel.

  11. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar


    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.


    Miller, G.


    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)


    Moore, W.T.


    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  14. Transients in reactors for power systems compensation (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  15. CHP Integrated with Burners for Packaged Boilers

    Energy Technology Data Exchange (ETDEWEB)

    Castaldini, Carlo; Darby, Eric


    The objective of this project was to engineer, design, fabricate, and field demonstrate a Boiler Burner Energy System Technology (BBEST) that integrates a low-cost, clean burning, gas-fired simple-cycle (unrecuperated) 100 kWe (net) microturbine (SCMT) with a new ultra low-NOx gas-fired burner (ULNB) into one compact Combined Heat and Power (CHP) product that can be retrofit on new and existing industrial and commercial boilers in place of conventional burners. The Scope of Work for this project was segmented into two principal phases: (Phase I) Hardware development, assembly and pre-test and (Phase II) Field installation and demonstration testing. Phase I was divided into five technical tasks (Task 2 to 6). These tasks covered the engineering, design, fabrication, testing and optimization of each key component of the CHP system principally, ULNB, SCMT, assembly BBEST CHP package, and integrated controls. Phase I work culminated with the laboratory testing of the completed BBEST assembly prior to shipment for field installation and demonstration. Phase II consisted of two remaining technical tasks (Task 7 and 8), which focused on the installation, startup, and field verification tests at a pre-selected industrial plant to document performance and attainment of all project objectives. Technical direction and administration was under the management of CMCE, Inc. Altex Technologies Corporation lead the design, assembly and testing of the system. Field demonstration was supported by Leva Energy, the commercialization firm founded by executives at CMCE and Altex. Leva Energy has applied for patent protection on the BBEST process under the trade name of Power Burner and holds the license for the burner currently used in the product. The commercial term Power Burner is used throughout this report to refer to the BBEST technology proposed for this project. The project was co-funded by the California Energy Commission and the Southern California Gas Company (SCG), a

  16. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.


    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  17. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter


    to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...... and malfunctions of anaerobic digesters occasionally experienced, leading to sub-optimal methane production and wastewater treatment. Using a variety of molecular techniques, we are able to determine which microorganisms are active, where they are active, and when they are active, but we still need to determine...

  18. Conceptual design of minor actinides burner with an accelerator-driven subcritical system.

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Y.; Gohar, Y. (Nuclear Engineering Division)


    In the environmental impact study of the Yucca Mountain nuclear waste repository, the limit of spent nuclear fuel (SNF) for disposal is assessed at 70,000 metric tons of heavy metal (MTHM), among which 63,000 MTHM are the projected SNF discharge from U.S. commercial nuclear power plants though 2011. Within the 70,000 MTHM of SNF in storage, approximately 115 tons would be minor actinides (MAs) and 585 tons would be plutonium. This study describes the conceptual design of an accelerator-driven subcritical (ADS) system intended to utilize (burn) the 115 tons of MAs. The ADS system consists of a subcritical fission blanket where the MAs fuel will be burned, a spallation neutron source to drive the fission blanket, and a radiation shield to reduce the radiation dose to an acceptable level. The spallation neutrons are generated from the interaction of a 1 GeV proton beam with a lead-bismuth eutectic (LBE) or liquid lead target. In this concept, the fission blanket consists of a liquid mobile fuel and the fuel carrier can be LBE, liquid lead, or molten salt. The actinide fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Therefore, fresh fuel can be fed into the fission blanket to adjust its reactivity and to control system power during operation. Monte Carlo analyses were performed to determine the overall parameters of an ADS system utilizing LBE as an example. Steady-state Monte Carlo simulations were studied for three fission blanket configurations that are similar except that the loaded amount of actinide fuel in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factor values of the three configurations are all approximately 0.98 and the MA initial inventories are each approximately 10 tons. Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. Preliminary burnup analysis shows that all three conceptual ADS

  19. Rodded shutdown system for a nuclear reactor (United States)

    Golden, Martin P.; Govi, Aldo R.


    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  20. Hybrid Molten Salt Reactor (HMSR) System Study

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL


    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  1. Control system studies for thermionic reactors (United States)

    Hermsen, R. J.; Gronroos, H. G.


    In core thermionic reactor concepts are of interest for space missions that require electrical power in the range of a few tens of kilowatts up to several megawatts. The physical principle involved--thermionic direct conversion of heat to electricity at net efficiencies up to 15 percent--offers potential advantages when compared to other nuclear powerplant concepts. However, the integration of the thermionic diode electrode structure with high-temperature nuclear fuel materials presents new design problems and new reactor physical constraints. Among the topics that must be investigated are those associated with the control system. The results of analytical and simulation studies of thermionic reactor control performed at the Jet Propulsion Laboratory are discussed.

  2. Reactor power system deployment and startup (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.


    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  3. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems. (United States)


    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  4. Industrial Energy Conservation, Forced Internal Recirculation Burner

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Rabovitser


    The overall objective of this research project is to develop and evaluate an industrial low NOx burner for existing and new gas-fired combustion systems for intermediate temperature (1400 degree to 2000 degree F) industrial heating devices such as watertube boilers and process fluid heaters. A multi-phase effort is being pursued with decision points to determine advisability of continuance. The current contract over Phases II and III of this work. The objectives of each phase are as follows. Phase II - to design, fabricate, and evaluate prototype burners based on the Forced Internal Recirculation (FIR) concept. Phase III - to evaluate the performance of an FIR burner under actual operating conditions in a full-scale field test and establish the performance necessary for subsequent commercialization

  5. Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system (United States)

    Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun


    In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

  6. Integral reactor system and method for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.


    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  7. Gas flat flame burner

    Energy Technology Data Exchange (ETDEWEB)

    Revun, M.P.; Chernov, V.Ye.; Perelman, L.D.; Rudnitskiy, O.I.; Yerinov, A.Ye.; Zyryanov, V.V.


    A burner is proposed in which it is possible to burn gas with low calorific value, Q/SUB n//SUP p/. The burner consists of a housing and screw-shaped insert installed on the central gassupply pipe. The latter ends at the outlet adapter equipped with nozzles of elliptical outlet section directed towards twisting of the air stream. The nozzles are bent at a right angle, and their axes are also arranged at the angle ..cap alpha.. in relation to the section plane of the adapter. ..cap alpha.. changes in limits of 0-45/sup 0/ depending on Q /SUB n/ /SUP p/ the lower the calorific value of the burned fuel, the higher the size of the angle ..cap alpha.. might be.

  8. Experimental investigation and optimisation of burner systems for glass melting ends with regenerative air preheating. Final report; Experimentelle Untersuchung und Optimierung von Brennersystemen fuer Glasschmelzwannen mit regenerativer Luftvorwaermung. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Scherello, A.; Flamme, M.; Kremer, H.


    The project comprised experiments on burner systems for glass melting ends with regenerative air preheating for the purpose of optimisation. The experimental set-up was to reflect realistic conditions. In the first stage of the investigations, modern burner systems were installed in a GWI test facility and investigated. [German] Ziel des oben genannten Forschungsvorhabens war die Durchfuehrung experimenteller Untersuchungen von Brennersystemen fuer Glasschmelzwannen mit regenerativer Luftvorwaermung sowie deren Optimierung. Dazu war es notwendig, einen experimentellen Aufbau zu realisieren, mit dessen Hilfe die Stroemungs-, Mischungs- und Umsetzungsphaenomene von Glasschmelzoefen realistisch nachgestellt und aussagekraeftige Untersuchungen durchgefuehrt werden koennen. In einem ersten Untersuchungsschritt wurden moderne Brennerlanzen an der GWI-Versuchsanlage installiert und untersucht. (orig.)

  9. Microchannel Reactor System for Catalytic Hydrogenation

    Energy Technology Data Exchange (ETDEWEB)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie


    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  10. Reactor Design for Bioelectrochemical Systems

    KAUST Repository

    Mohanakrishna, G.


    Bioelectrochemical systems (BES) are novel hybrid systems which are designed to generate renewable energy from the low cost substrate in a sustainable way. Microbial fuel cells (MFCs) are the well studied application of BES systems that generate electricity from the wide variety of organic components and wastewaters. MFC mechanism deals with the microbial oxidation of organic molecules for the production of electrons and protons. The MFC design helps to build the electrochemical gradient on anode and cathode which leads for the bioelectricity generation. As whole reactions of MFCs happen at mild environmental and operating conditions and using waste organics as the substrate, it is defined as the sustainable and alternative option for global energy needs and attracted worldwide researchers into this research area. Apart from MFC, BES has other applications such as microbial electrolysis cells (MECs) for biohydrogen production, microbial desalinations cells (MDCs) for water desalination, and microbial electrosynthesis cells (MEC) for value added products formation. All these applications are designed to perform efficiently under mild operational conditions. Specific strains of bacteria or specifically enriched microbial consortia are acting as the biocatalyst for the oxidation and reduction of BES. Detailed function of the biocatalyst has been discussed in the other chapters of this book.

  11. Laser fusion hybrid reactor systems study

    Energy Technology Data Exchange (ETDEWEB)


    The work was performed in three phases. The first phase included a review of the many possible laser-reactor-blanket combinations and resulted in the selection of a ''demonstration size'' 500 MWe plant for further study. A number of fast fission blankets using uranium metal, uranium-molybdenum alloy, and uranium carbide as fuel were investigated. The second phase included design of the reactor vessel and internals, heat transfer system, tritium processing system, and the balance of plant, excluding the laser building and equipment. A fuel management scheme was developed, safety considerations were reviewed, and capital and operating costs were estimated. Costs developed during the second phase were unexpectedly high, and a thorough review indicated considerable unit cost savings could be obtained by scaling the plant to a larger size. Accordingly, a third phase was added to the original scope, encompassing the redesign and scaling of the plant from 500 MWe to 1200 MWe (less lasers).

  12. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system (United States)

    Harto, Andang Widi


    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  13. Power conditioning for space nuclear reactor systems (United States)

    Berman, Baruch


    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  14. Staged membrane oxidation reactor system (United States)

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh


    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  15. Nuclear reactor pressure vessel support system (United States)

    Sepelak, George R.


    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,


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

  17. Weld monitor and failure detector for nuclear reactor system (United States)

    Sutton, Jr., Harry G.


    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  18. Burner Characteristics for Activated Carbon Production

    Directory of Open Access Journals (Sweden)

    zakaria Supaat


    Full Text Available Carbonization process has become an important stage in developing activated carbon. However, existing burner are not efficient in time production which take 24 hours to15 days for charcoal production. Therefore, new design of burner/kilns is quite needed in order to produce larger number of charcoal in short time production, to improve charcoal quality regarding to the smooth surface area and pore volume. This research proposed new design burner which divided into two types which are vertical and horizontal types. Vertical is not completed by auto-rotating system while horizontal type is complete by auto-rotating and fume handling system. It developed using several equipment such as welding, oxy-cutting, drilling grinding and cutting machine. From the result of carbonization process shows that coconut shell charcoal need shorter time of 30 minutes as compared to palm shell charcoal of 2 h to completely carbonized. This result claim that the new design better than existing kiln that need longer time up to 24 h. The result of the palm and coconut shell charcoal believe will produce better properties of activated carbon in large surface area and higher total volume of pores. Therefore, this burner is high recommended for producing palm and coconut shell charcoal as well as other bio-based material.


    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski


    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  20. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D


    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

  1. Integrated systems analysis of the PIUS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others


    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  2. Catalyzed Ceramic Burner Material

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, Amy S., Dr.


    Catalyzed combustion offers the advantages of increased fuel efficiency, decreased emissions (both NOx and CO), and an expanded operating range. These performance improvements are related to the ability of the catalyst to stabilize a flame at or within the burner media and to combust fuel at much lower temperatures. This technology has a diverse set of applications in industrial and commercial heating, including boilers for the paper, food and chemical industries. However, wide spread adoption of catalyzed combustion has been limited by the high cost of precious metals needed for the catalyst materials. The primary objective of this project was the development of an innovative catalyzed burner media for commercial and small industrial boiler applications that drastically reduce the unit cost of the catalyzed media without sacrificing the benefits associated with catalyzed combustion. The scope of this program was to identify both the optimum substrate material as well as the best performing catalyst construction to meet or exceed industry standards for durability, cost, energy efficiency, and emissions. It was anticipated that commercial implementation of this technology would result in significant energy savings and reduced emissions. Based on demonstrated achievements, there is a potential to reduce NOx emissions by 40,000 TPY and natural gas consumption by 8.9 TBtu in industries that heavily utilize natural gas for process heating. These industries include food manufacturing, polymer processing, and pulp and paper manufacturing. Initial evaluation of commercial solutions and upcoming EPA regulations suggests that small to midsized boilers in industrial and commercial markets could possibly see the greatest benefit from this technology. While out of scope for the current program, an extension of this technology could also be applied to catalytic oxidation for volatile organic compounds (VOCs). Considerable progress has been made over the course of the grant

  3. Lead-cooled fast reactor (LFR) overview and perspectives


    CINOTTI Luciano; Smith, Craig F.; SEKIMOTO, HIROSHI


    The GIF Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology with great potential to meet the small-unit electricity needs of remote sites while also offering advantages as a large system for grid-connected power stations. The LFR features a fast- neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner of minor actinides from spent fuel and as a burner/breeder. An important feature of the LFR is the ...

  4. Refinery burner simulation design architecture summary.

    Energy Technology Data Exchange (ETDEWEB)

    Pollock, Guylaine M.; McDonald, Michael James; Halbgewachs, Ronald D.


    This report describes the architectural design for a high fidelity simulation of a refinery and refinery burner, including demonstrations of impacts to the refinery if errors occur during the refinery process. The refinery burner model and simulation are a part of the capabilities within the Sandia National Laboratories Virtual Control System Environment (VCSE). Three components comprise the simulation: HMIs developed with commercial SCADA software, a PLC controller, and visualization software. All of these components run on different machines. This design, documented after the simulation development, incorporates aspects not traditionally seen in an architectural design, but that were utilized in this particular demonstration development. Key to the success of this model development and presented in this report are the concepts of the multiple aspects of model design and development that must be considered to capture the necessary model representation fidelity of the physical systems.

  5. Catalytic burners in larger boiler appliances

    Energy Technology Data Exchange (ETDEWEB)

    Silversand, Fredrik; Persson, Mikael (Catator AB, Lund (Sweden))


    boiler design to benefit from all advantages provided by catalytic combustion. Such work will include design and construction of an innovative boiler system, in the effect segment of 100- 200 kWt. The boiler system will then be evaluated with Catator's catalytic burner technology

  6. Mechanical systems development of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Chang, M. H.; Kim, J. I.; Choi, S.; Kim, K. S.; Kim, T. W.; Jeong, K. H.; Kim, J. H.; Kim, Y. W.; Lee, G. M.


    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose applications such as small capacity power generation, co-generation and sea water desalination. This in mind, survey has been made on the worldwide small and medium integral reactors under development. Reviewed are their technical characteristics, development status, design features, application plans, etc. For the mechanical design scope of work, the structural concept compatible with the characteristics and requirements of integral reactor has been established. Types of major components were evaluated and selected. Functional and structural concept, equipment layout and supporting concept within the reactor pressure vessel have also been established. Preliminary mechanical design requirements were developed considering the reactor lifetime, operation conditions, and the expected loading combinations. To embody the concurrent design approach, recent CAD technology and team engineering concept were evaluated. (author). 31 refs.,16 tabs., 35 figs.

  7. DNA-Based Enzyme Reactors and Systems

    Directory of Open Access Journals (Sweden)

    Veikko Linko


    Full Text Available During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications.

  8. Systems aspects of a space nuclear reactor power system (United States)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.


    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  9. High Performance Photocatalytic Oxidation Reactor System Project (United States)

    National Aeronautics and Space Administration — Pioneer Astronautics proposes a technology program for the development of an innovative photocatalytic oxidation reactor for the removal and mineralization of...

  10. Deposit formation by 20 % (V/V) FAME fuels in premix burner systems; Ablagerungsbildung durch 20% (V/V) FAME-Brennstoffe in Vormischbrennersystemen

    Energy Technology Data Exchange (ETDEWEB)

    Jaschinski, Christian; Rheinberg, Oliver van [OWI Oel-Waerme-Institut GmbH, Aachen (Germany); RWTH Aachen (Germany). An-Institut


    In the domestic heating market the development and use of fuels with an increasing share of biogenic or alternative fuels is propagated. Due to the fact, that modern fuel oil burner feature a complex carburation techniques and combustion, changes on the fuel properties and composition can lead to increased emissions or deposit formation therein. Furthermore, the different fuel properties may result in decreased storage stability, which has to be evaluated before introducing them into the market. The scope of the project was to investigate the performance of low-sulfur domestic heating oil (DHO) with up to 20 % v/v FAME on the storage stability and on the use in oil-fired heating systems. The project was split into two major parts. The first part covered a two-year storage of the fuels including sampling and analysis of the fuels every half year. The analysis was conducted according to DIN 51603-1 for the pure DHO and according to DIN SPEC 51603-6 for the blends. It has been shown, that low sulphur domestic heating oil with up to 20 % (V/V) of FAME after two years of storage fits the parameter of the corresponding standards. Furthermore, a new testing method, called 'DGMK-714' derived from the PetroOxy-test (EN 16091) has been defined. With this method for the determination of oxidation stability the fuels can be characterized being comparable to the standardized testing methods of modified Rancimat or PetroOxy. The higher sample volume of the method allows further analysis of the fuel sample after testing for characterization of the fuels. The second part of the project investigated the deposit formation tendencies of the fuels in an idealized testing apparatus and in three different kinds of oil burners. Using the idealized testing apparatus proved an increased tendency of deposit formation during evaporation for an increasing FAME content. However, this tendency could not be observed in the three commercial oil-fired heating systems. A precise fuel

  11. REACTOR - a Concept for establishing a System-of-Systems (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim


    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

  12. An analysis system for in-reactor behavior, FANTASI

    Energy Technology Data Exchange (ETDEWEB)

    Uto, Nariaki; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sugaya, Toshio; Sakai, Kimiaki [Japan Nucler Cycle Developmnet Inst., Tokai, Ibaraki (Japan)


    The Japan Nuclear Fuel Cycle Development Institute developed FANTASI (A Computational System For Analyzing Coupled Neutronic, Thermal-Hydraulic And Structural Behaviors In A Fast Breeder Reactor Core) to simulate a conditions where nuclear reaction, thermal-hydraulic behavior of coolant and deformation of core construction progress under mutual relation in reactor of a fast breeder reactor by cooperation of engineers in the fields of physics, thermal-hydraulics, structure, and information system on reactor. Here was described on system construction of FANTASI after describing progress of this development. And then, after introducing a case study using this system, applicability to transient phenomena in nuclear reactor was described. At last, with summarizing results of this development, its future development was also mentioned. (G.K.)

  13. Passive modular gas safety system for a reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Isaev, I.F.; Kulakov, A.A.; Sivokon, V.P.; Udovenko, A.N.; Ionaitis, R.R.


    Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

  14. Research, development, and testing of a prototype two-stage low-input rate oil burner for variable output heating system applications

    Energy Technology Data Exchange (ETDEWEB)

    Krajewski, R.F.; Butcher, T.A. [Brookhaven National Labs., Upton, NY (United States)


    The use of a Two-Stage Fan Atomized Oil Burner (TSFAB) in space and water heating applications will have dramatic advantages in terms of it`s potential for a high Annual Fuel Utilization Efficiency (AFUE) and/or Energy Factor (EF) rating for the equipment. While demonstrations of a single rate burner in an actual application have already yielded sufficient confidence that space and domestic heating loads can be met at a single low firing rate, this represents only a narrow solution to the diverse nature of building space heating and domestic water loads that the industry must address. The mechanical development, proposed control, and testing of the Two-Stage burner is discussed in terms of near term and long term goals.

  15. SP-100 Program: space reactor system and subsystem investigations

    Energy Technology Data Exchange (ETDEWEB)

    Harty, R.B.


    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  16. A new VFA sensor technique for anaerobic reactor systems

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær


    to monitor VFA online in one of the most difficult media: animal slurry or manure. A novel in situ filtration technique has made it possible to perform microfiltration inside a reactor system. This filter enables sampling from closed reactor systems without large-scale pumping and filters. Furthermore, due...... to its small size it can be placed in lab-scale reactors without disturbing the process. Using this filtration technique together with commercially available membrane filters we have constructed a VFA sensor system that can perform automatic analysis of animal slurry at a frequency as high as every 15...... filtration technique are being presented is this article....

  17. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems (United States)

    Was, Gary S.


    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  18. Design and analysis of the federal aviation administration next generation fire test burner (United States)

    Ochs, Robert Ian

    The United States Federal Aviation Administration makes use of threat-based fire test methods for the certification of aircraft cabin materials to enhance the level of safety in the event of an in-flight or post-crash fire on a transport airplane. The global nature of the aviation industry results in these test methods being performed at hundreds of laboratories around the world; in some cases testing identical materials at multiple labs but yielding different results. Maintenance of this standard for an elevated level of safety requires that the test methods be as well defined as possible, necessitating a comprehensive understanding of critical test method parameters. The tests have evolved from simple Bunsen burner material tests to larger, more complicated apparatuses, requiring greater understanding of the device for proper application. The FAA specifies a modified home heating oil burner to simulate the effects of large, intense fires for testing of aircraft seat cushions, cargo compartment liners, power plant components, and thermal acoustic insulation. Recently, the FAA has developed a Next Generation (NexGen) Fire Test burner to replace the original oil burner that has become commercially unavailable. The NexGen burner design is based on the original oil burner but with more precise control of the air and fuel flow rates with the addition of a sonic nozzle and a pressurized fuel system. Knowledge of the fundamental flow properties created by various burner configurations is desired to develop an updated and standardized burner configuration for use around the world for aircraft materials fire testing and airplane certification. To that end, the NexGen fire test burner was analyzed with Particle Image Velocimetry (PIV) to resolve the non-reacting exit flow field and determine the influence of the configuration of burner components. The correlation between the measured flow fields and the standard burner performance metrics of flame temperature and

  19. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System (United States)

    Tournier, Jean-Michel; El-Genk, Mohamed S.


    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  20. Numerical simulation of porous burners and hole plate surface burners

    Directory of Open Access Journals (Sweden)

    Nemoda Stevan


    Full Text Available In comparison to the free flame burners the porous medium burners, especially those with flame stabilization within the porous material, are characterized by a reduction of the combustion zone temperatures and high combustion efficiency, so that emissions of pollutants are minimized. In the paper the finite-volume numerical tool for calculations of the non-isothermal laminar steady-state flow, with chemical reactions in laminar gas flow as well as within porous media is presented. For the porous regions the momentum and energy equations have appropriate corrections. In the momentum equations for the porous region an additional pressure drop has to be considered, which depends on the properties of the porous medium. For the heat transfer within the porous matrix description a heterogeneous model is considered. It treats the solid and gas phase separately, but the phases are coupled via a convective heat exchange term. For the modeling of the reaction of the methane laminar combustion the chemical reaction scheme with 164 reactions and 20 chemical species was used. The proposed numerical tool is applied for the analyses of the combustion and heat transfer processes which take place in porous and surface burners. The numerical experiments are accomplished for different powers of the porous and surface burners, as well as for different heat conductivity character is tics of the porous regions.

  1. Emergency heat removal system for a nuclear reactor (United States)

    Dunckel, Thomas L.


    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  2. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)


    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  3. Autonomous Control of Space Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na


    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  4. A Design of Alarm System in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants.

  5. Digital, remote control system for a 2-MW research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Battle, R.E.; Corbett, G.K.


    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  6. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.


    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  7. Deployment history and design considerations for space reactor power systems (United States)

    El-Genk, Mohamed S.


    The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

  8. New reactor technology: safety improvements in nuclear power systems. (United States)

    Corradini, M L


    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  9. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, C.W.; Stevens, W.W.


    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

  10. Modeling and simulation of CANDU reactor and its regulating system (United States)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  11. Design of virtual SCADA simulation system for pressurized water reactor (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman


    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  12. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail:; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)


    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  13. Simplified safety and containment systems for the iris reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.E. [Westinghouse Electric Co., Pittsburgh, PA (United States); Lombardi, C.; Ricotti, M.; Oriani, L. [Polytechnic of Milan, Dept. of Nuclear Engineering, Milan (Italy)


    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  14. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    Energy Technology Data Exchange (ETDEWEB)

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.


    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  15. Development of a High Fidelity System Analysis Code for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Vincent Mousseau; Haihua Zhao


    well as some results from analyzing a simplified primary system model of GNEP’s advanced burner test reactor (ABTR) designed by Argonne. Various transient analyses are performed with this simplified ABTR model to study two fundamental issues related to system analysis codes – accuracy of numeric algorithm and efficiency. The accuracy study is carried by comparing the second order method with the first order method. The results show that numerical errors in the first order method are large and it is very difficult to distinguish numerical errors from physical modeling errors. On the other hand, second order method yields small numerical errors and it is very easy to spot physical modeling errors. The efficiency study is carried out by comparing the time steps for the fully implicit solution algorithm versus CFL stability limit methods. The dynamic time steps used in a fully implicit method will adjust the time step to resolve the time scale during the various stages of a long lasting transient. This will make a computer code based on fully implicit methods run more efficiently versus a CFL stability limit method code like RELAP, in which a particle of fluid cannot cross a control volume in a single time step.

  16. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)


    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  17. Industrial Medium-Btu Fuel Gas Demonstration-Plant Program. Technical support report: combustion system data. Part 2. Burner conversion survey

    Energy Technology Data Exchange (ETDEWEB)



    This study was limited to an analysis of the feasibility of burning the IFG in the existing burners and combustion chambers among a group of prospective IFG customers. The results of this study indicate that the great majority of burner and equipment manufacturers recommend that the IFG can be utilized with their equipment. This is especially true with the boilers which make up the largest part of the load among the potential users of the IFG. A small number of burners representing a small part of the total potential load will probably have to be replaced. This study did not address the changes that would be required with respect to the fuel distribution piping within each facility. At a minimum of the existing regulators, flow meters, and control valves designed for the natural gas flow rates would have to be replaced to accommodate the higher fuel flow rates requiring with the IFG. In many facilities, the fuel distribution piping would have to be replaced. No changes, however, are requied for the combustion air fans or flues and stacks.

  18. Numerical investigation of a novel burner to combust anode exhaust gases of SOFC stacks

    Directory of Open Access Journals (Sweden)

    Pianko-Oprych Paulina


    Full Text Available The aim of the present study was a numerical investigation of the efficiency of the combustion process of a novel concept burner under different operating conditions. The design of the burner was a part of the development process of a complete SOFC based system and a challenging combination of technical requirements to be fulfilled. A Computational Fluid Dynamics model of a non-premixed burner was used to simulate combustion of exhaust gases from the anode region of Solid Oxide Fuel Cell stacks. The species concentrations of the exhaust gases were compared with experimental data and a satisfactory agreement of the conversion of hydrocarbons was obtained. This validates the numerical methodology and also proves applicability of the developed approach that quantitatively characterized the interaction between the exhaust gases and burner geometry for proper combustion modelling. Thus, the proposed CFD approach can be safely used for further numerical optimisation of the burner design.

  19. Pulverized fuel-oxygen burner (United States)

    Taylor, Curtis; Patterson, Brad; Perdue, Jayson


    A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through the solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.

  20. Optimization of burners in oxygen-gas fired glass furnace

    NARCIS (Netherlands)

    Kersbergen, M.J. van; Beerkens, R.G.C.; Sarmiento-Darkin, W.; Kobayashi, H.


    The energy efficiency performance, production stability and emissions of oxygen-fired glass furnaces are influenced by the type of burner, burner nozzle sizes, burner positions, burner settings, oxygen-gas ratios and the fuel distribution among all the burners. These parameters have been optimized

  1. Small reactor power systems for manned planetary surface bases (United States)

    Bloomfield, Harvey S.


    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  2. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  3. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  4. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  5. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)


    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  6. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri


    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  7. System and method for temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.


    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  8. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)


    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  9. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)


    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  10. Fault detection system for Argentine Research Reactor instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Polenta, H.P. (Argentine Navy, Comodoro Py 2055 Office 11-93, 1104 - Buenos Aires (Argentina)); Bernard, J.A. (Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany Street, Cambridge, Massachusetts 02139 (United States)); Ray, A. (205 Mechanical Engineering Department, Pennsylvania State University, University Park, Pennsylvania 16802 (United States))


    The design and implementation of a redundancy management scheme for the on-line detection and isolation of faulty sensors is presented. Such a device is potentially useful in reactor-powered spacecraft for enhancing the processing capabilities of the main computer. The fault detection device can be used as an integral part of intelligent instrumentation systems. The device has been built using an 8-bit microcontroller and commercially available electronic hardware. The software is completely portable. The operation of this device has been successfully demonstrated for real-time validation of sensor data on Argentina's RA-1 Research Reactor.

  11. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.


    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...... extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between...

  12. Space-reactor electric systems: subsystem technology assessment

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.V.; Bost, D.; Determan, W.R.


    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  13. Modification of the Core Cooling System of TRIGA 2000 Reactor (United States)

    Umar, Efrizon; Fiantini, Rosalina


    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  14. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others


    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  15. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

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


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

  16. Computational fluid dynamics in oil burner design

    Energy Technology Data Exchange (ETDEWEB)

    Butcher, T.A. [Brookhaven National Labs., Upton, NY (United States)


    In Computational Fluid Dynamics, the differential equations which describe flow, heat transfer, and mass transfer are approximately solved using a very laborious numerical procedure. Flows of practical interest to burner designs are always turbulent, adding to the complexity of requiring a turbulence model. This paper presents a model for burner design.

  17. Study of Natural Convection Passive Cooling System for Nuclear Reactors (United States)

    Abdillah, Habibi; Saputra, Geby; Novitrian; Permana, Sidik


    Fukushima nuclear reactor accident occurred due to the reactor cooling pumps and followed by all emergencies cooling systems could not work. Therefore, the system which has a passive safety system that rely on natural laws such as natural convection passive cooling system. In natural convection, the cooling material can flow due to the different density of the material due to the temperature difference. To analyze such investigation, a simple apparatus was set up and explains the study of natural convection in a vertical closed-loop system. It was set up that, in the closed loop, there is a heater at the bottom which is representing heat source system from the reactor core and cooler at the top which is showing the cooling system performance in room temperature to make a temperature difference for convection process. The study aims to find some loop configurations and some natural convection performances that can produce an optimum flow of cooling process. The study was done and focused on experimental approach and simulation. The obtained results are showing and analyzing in temperature profile data and the speed of coolant flow at some point on the closed-loop system.

  18. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Ahmmed Saadi Ibrehem


    Full Text Available A modified model for the neutralization process of Stirred Tank Reactors (CSTR reactor is presented in this study. The model accounts for the effect of strong acid [HCL] flowrate and strong base [NaOH] flowrate with the ionic concentrations of [Cl-] and [Na+] on the Ph of the system. In this work, the effect of important reactor parameters such as ionic concentrations and acid and base flowrates on the dynamic behavior of the CSTR is investigated and the behavior of mathematical model is compared with the reported models for the McAvoy model and Jutila model. Moreover, the results of the model are compared with the experimental data in terms of pH dynamic study. A good agreement is observed between our model prediction and the actual plant data. © 2011 BCREC UNDIP. All rights reserved(Received: 1st March 2011, Revised: 28th March 2011; Accepted: 7th April 2011[How to Cite: A.S. Ibrehem. (2011. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 6(1: 47-52. doi:10.9767/bcrec.6.1.825.47-52][How to Link / DOI: || or local: ] | View in 

  19. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)


    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  20. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.


    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  1. Operation of staged membrane oxidation reactor systems (United States)

    Repasky, John Michael


    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  2. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors (United States)


    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  3. Burners (United States)

    ... and Sugar Substitutes Exercise and Fitness Exercise Basics Sports Safety Injury Rehabilitation Emotional Well-Being Mental Health Sex and ... and Sugar Substitutes Exercise and Fitness Exercise Basics Sports Safety Injury Rehabilitation Emotional Well-Being Mental Health Sex and ...

  4. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.


    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  5. Monitoring system for a liquid-cooled nuclear fission reactor (United States)

    DeVolpi, Alexander


    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.


    Directory of Open Access Journals (Sweden)

    M.M. Noor


    Full Text Available This paper discusses the design and development of the Moderate and Intense Low oxygen Dilution (MILD combustion burner using Computational Fluid Dynamics (CFD simulations. The CFD commercial package was used to simulate preliminary designs for the burner before the final design was sent to the workshop for fabrication. The burner is required to be a non-premixed and open burner. To capture and use the exhaust gas, the burner was enclosed within a large circular shaped wall with an opening at the top. An external EGR pipe was used to transport the exhaust gas which was mixed with the fresh oxidant. To control the EGR and exhaust flow, butterfly valves were installed at the top opening as a damper to close the exhaust gas flow at a certain ratio for EGR and exhaust out to the atmosphere. High temperature fused silica glass windows were installed to view and capture images of the flame and analyze the flame propagation. The burner simulation shows that MILD combustion was achieved for the oxygen mole fraction of 3-13%. The final design of the burner was fabricated and ready for the experimental validation.

  7. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)


    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  8. Systems and methods for dismantling a nuclear reactor (United States)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon


    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  9. Development of fluid system design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kang, D. J. and others


    This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

  10. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.


    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  11. Towards a better understanding of biomass suspension co-firing impacts via investigating a coal flame and a biomass flame in a swirl-stabilized burner flow reactor under same conditions

    DEFF Research Database (Denmark)

    Yin, Chungen; Rosendahl, Lasse; Kær, Søren Knudsen


    increases the residence time of coal particles. Both the factors favor a complete burnout of the coal particles. The higher volatile yields of the straw produce more off-gas, requiring more O2 for the fast gas phase combustion and causing the off-gas to proceed to a much larger volume in the reactor prior...... to mixing with oxidizer. For the pulverized straw particles of a few hundred microns in diameters, the intra-particle conversion is found to be a secondary issue at most in their combustion. The simulations also show that a simple switch of the straw injection mode can not improve the burnout of the straw...

  12. Development of ROV System for FOSAR in Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Soo; Kim, Tae Won; Lee, Sung Uk; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Nam Kyun [Korea Plant Service and Engineering Co., Seongnam (Korea, Republic of)


    Foreign object in the reactor vessel is susceptible to damage the fuel. Prior to reloading fuel assemblies into the core, FOSAR(Foreign Object Search And Retrieval) activities were performed on and beneath the lower core plate with conventional equipment. However, the reactor vessel is limited to humans who are susceptible to radiation exposure, and conventional equipment is hard to access because of the complexity of the structure. To improve the convenience of use and retrieval ability in the under-core plate region, we are developing a FOSAR system carried by ROV (Remotely Operated Vehicle). In this paper, we describe a ROV system developed. The ROV system is composed of robot vehicle and remote control unit. The vehicle has 4 thrusters, tilt, camera, light and depth sensor, etc. Considering radiation damage, processors are not equipped on the vehicle. Control signals and sensing signals are transferred through umbilical cable. Remote control unit is composed of electric driving module and two computers which one is for the control and the other is for the detection of robot position. Control computer has a joystick user input and video/signal input, and transmit motor control signal and lens control signal via CAN/RS485 communication. And the other computers transmit information of vehicle position to the control computer via serial communication. Information of vehicle position is obtained through image processing algorithm. The acquiring camera of vehicle is on the flange of reactor vessel. Simulations on the detection of vehicle position are performed at the reactor vessel mockup which scaled down by 6 and verified to use in the control of robot by visual tracking. And functional test has been performed on the air condition. In the future, performance test will be carried out real sized mockup and underwater condition

  13. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)


    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  14. 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


    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


    Energy Technology Data Exchange (ETDEWEB)

    E.G. Eddings; A. Molina; D.W. Pershing; A.F. Sarofim; T.H. Fletcher; H. Zhang; K.A. Davis; M. Denison; H. Shim


    The focus of this program is to provide insight into the formation and minimization of NO{sub x} in multi-burner arrays, such as those that would be found in a typical utility boiler. Most detailed studies are performed in single-burner test facilities, and may not capture significant burner-to-burner interactions that could influence NO{sub x} emissions. Thus, investigations of such interactions were made by performing a combination of single and multiple burner experiments in a pilot-scale coal-fired test facility at the University of Utah, and by the use of computational combustion simulations to evaluate full-scale utility boilers. In addition, fundamental studies on nitrogen release from coal were performed to develop greater understanding of the physical processes that control NO formation in pulverized coal flames--particularly under low NO{sub x} conditions. A CO/H{sub 2}/O{sub 2}/N{sub 2} flame was operated under fuel-rich conditions in a flat flame reactor to provide a high temperature, oxygen-free post-flame environment to study secondary reactions of coal volatiles. Effects of temperature, residence time and coal rank on nitrogen evolution and soot formation were examined. Elemental compositions of the char, tar and soot were determined by elemental analysis, gas species distributions were determined using FTIR, and the chemical structure of the tar and soot was analyzed by solid-state {sup 13}C NMR spectroscopy. A laminar flow drop tube furnace was used to study char nitrogen conversion to NO. The experimental evidence and simulation results indicated that some of the nitrogen present in the char is converted to nitric oxide after direct attack of oxygen on the particle, while another portion of the nitrogen, present in more labile functionalities, is released as HCN and further reacts in the bulk gas. The reaction of HCN with NO in the bulk gas has a strong influence on the overall conversion of char-nitrogen to nitric oxide; therefore, any model that

  16. Automated power control system for reactor TRIGA PUSPATI (United States)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair


    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  17. District heating biofuel burner efficiency and energy balance


    Okoro, Oluwashola Aderemi


    District heating is an optimal system of distributing heat to residential building in a centralized location through pipeline networks. The district heating of woodchip is cost effective, improve energy efficiency, reduce gas emissions and improve energy security. The thermal efficiency and energy balance in a boiler is obtained by combustion analysis of the wood (fuel). In this report, the district heating bio fuel burner in Skien Fjernvarme is considered. The capacity of the boiler is 6MW a...

  18. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Kuscu, Ozlem Selcuk [Department of Environmental Engineering, Faculty of Engineering and Architecture, Sueleyman Demirel University, 32360, Isparta (Turkey); Sponza, Delia Teresa, E-mail: [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, 35160, Izmir (Turkey)


    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m{sup 3} day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m{sup 3} day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m{sup 3} day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m{sup 3} day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m{sup 3} day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m{sup 3} day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from

  19. Ongoing Development of a Series Bosch Reactor System (United States)

    Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David


    Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

  20. The detector system of the Daya Bay reactor neutrino experiment (United States)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.


    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  1. TCODE: a computer code for analysis of tritium and vacuum systems for tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.


    TCODE can be used for either near-term experimental reactors or for commercial reactors. The code provides options for items that may be included in a commercial reactor such as a divertor, neutral beam heating, and a breeding blanket. The code was used to calculate tritium and vacuum system parameters for the near term reactors ITR, TNS-UP and EPR as well as for some commercial reactor designs, the UWMAK series. A selected sample of the tritium and vacuum parameters for these reactor designs is shown. Also shown are parameters for a hypothetical reactor UWMAK-III M having similar characteristics to UWMAK-III but with a higher fractional burnup (5.0% cf. 0.83%). The impact of the reactor design scenario upon major tritium and vacuum systems is discussed.

  2. Anaerobic sewage treatment in a one-stage UASB reactor and a combined UASB Digester system

    NARCIS (Netherlands)

    Mahmoud, N.A.; Zeeman, G.; Gijzen, H.J.; Lettinga, G.


    The treatment of sewage at 15°C was investigated in a one-stage upflow anaerobic sludge blanket (UASB) reactor and a UASB-Digester system. The latter consists of a UASB reactor complemented with a digester for mutual sewage treatment and sludge stabilisation. The UASB reactor was operated at a

  3. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)


    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  4. Burners and combustion apparatus for carbon nanomaterial production (United States)

    Alford, J. Michael; Diener, Michael D; Nabity, James; Karpuk, Michael


    The invention provides improved burners, combustion apparatus, and methods for carbon nanomaterial production. The burners of the invention provide sooting flames of fuel and oxidizing gases. The condensable products of combustion produced by the burners of this invention produce carbon nanomaterials including without limitation, soot, fullerenic soot, and fullerenes. The burners of the invention do not require premixing of the fuel and oxidizing gases and are suitable for use with low vapor pressure fuels such as those containing substantial amounts of polyaromatic hydrocarbons. The burners of the invention can operate with a hot (e.g., uncooled) burner surface and require little, if any, cooling or other forms of heat sinking. The burners of the invention comprise one or more refractory elements forming the outlet of the burner at which a flame can be established. The burners of the invention provide for improved flame stability, can be employed with a wider range of fuel/oxidizer (e.g., air) ratios and a wider range of gas velocities, and are generally more efficient than burners using water-cooled metal burner plates. The burners of the invention can also be operated to reduce the formation of undesirable soot deposits on the burner and on surfaces downstream of the burner.

  5. Designing a SCADA system simulator for fast breeder reactor (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.


    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  6. Parametric systems analysis of the Modular Stellarator Reactor (MSR)

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.


    The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas.

  7. Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

    Directory of Open Access Journals (Sweden)

    Isaac Skavdahl


    Full Text Available Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco and the hot outlet temperature of the intermediate heat exchanger (Tho2 by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2 responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1 flow rate manipulation; (2 reactor power manipulation; or (3 a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  8. Catalytic membrane reactor for tritium extraction system from He purge

    Energy Technology Data Exchange (ETDEWEB)

    Santucci, Alessia, E-mail: [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Incelli, Marco [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); DEIM, University of Tuscia, Via del Paradiso 47, 01100 Viterbo (Italy); Sansovini, Mirko; Tosti, Silvano [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy)


    Highlights: • In the HCBB blanket, the produced tritium is recovered by purging with helium; membrane technologies are able to separate tritium from helium. • The paper presents the results of two experimental campaigns. • In the first, a Pd–Ag diffuser for hydrogen separation is tested at several operating conditions. • In the second, the ability of a Pd–Ag membrane reactor for water decontamination is assessed by performing isotopic swamping and water gas shift reactions. - Abstract: In the Helium Cooled Pebble Bed (HCPB) blanket concept, the produced tritium is recovered purging the breeder with helium at low pressure, thus a tritium extraction system (TES) is foreseen to separate the produced tritium (which contains impurities like water) from the helium gas purge. Several R&D activities are running in parallel to experimentally identify most promising TES technologies: particularly, Pd-based membrane reactors (MR) are under investigation because of their large hydrogen selectivity, continuous operation capability, reliability and compactness. The construction and operation under DEMO relevant conditions (that presently foresee a He purge flow rate of about 10,000 Nm{sup 3}/h and a H{sub 2}/He ratio of 0.1%) of a medium scale MR is scheduled for next year, while presently preliminary experiments on a small scale reactor are performed to identify most suitable operative conditions and catalyst materials. This work presents the results of an experimental campaign carried out on a Pd-based membrane aimed at measuring the capability of this device in separating hydrogen from the helium. Many operative conditions have been investigated by considering different He/H{sub 2} feed flow ratios, several lumen pressures and reactor temperatures. Moreover, the performances of a membrane reactor (composed of a Pd–Ag tube having a wall thickness of about 113 μm, length 500 mm and diameter 10 mm) in processing the water contained in the purge gas have been

  9. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B.F.; Berry, D.T.


    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters.

  10. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)


    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  11. Improved reactor regulating system logical architecture using genetic algorithm

    Directory of Open Access Journals (Sweden)

    Hyo-Sub Shim


    Full Text Available An improved Reactor Regulating System (RRS logic architecture, which is combined with genetic algorithm (GA, is implemented in this work. It is devised to provide an optimal solution to the current RRS. The current system works desirably and has contributed to safe and stable nuclear power plant operation. However, during the ascent and descent section of the reactor power, the RRS output reveals a relatively high steady-state error, and the output also carries a considerable level of overshoot. In an attempt to consolidate conservatism and minimize the error, this work proposes to apply GA to RRS and suggests reconfiguring the system. Prior to the use of GA, reverse engineering is implemented to build a Simulink-based RRS model. Reengineering is followed to produce a newly configured RRS to generate an output that has a reduced steady-state error and diminished overshoot level. A full-scope APR1400 simulator is used to examine the dynamic behaviors of RRS and to build the RRS Simulink model.

  12. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)


    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  13. Development of system integration technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kang, D. J.; Kim, K. K. and others


    The objective of this report is to integrate the conceptual design of an integral reactor, SMART producing thermal energy of 330 MW, which will be utilized to supply energy for seawater desalination and small-scale power generation. This project also aims to develop system integration technology for effective design of the reactor. For the conceptual design of SMART, preliminary design requirements including the top-tier requirements and design bases were evaluated and established. Furthermore, in the view of the application of codes and standards to the SMART design, existing laws, codes and standards were analyzed and evaluated with respect to its applicability. As a part of this evaluation, directions and guidelines were proposed for the development of new codes and standards which shall be applied to the SMART design. Regarding the integration of SMART conceptual designs, major design activities and interfaces between design departments were established and coordinated through the design process. For the effective management of all design schedules, a work performance evaluation system was developed and applied to the design process. As the results of this activity, an integrated output of SMART designs was produced. Two additional scopes performed in this project include the preliminary economic analysis on the SMART utilization for seawater desalination, and the planning of verification tests for technology implemented into SMART and establishing development plan of the computer codes to be used for SMART design in the next phase. The technical cooperation with foreign country and international organization for securing technologies for integral reactor design and its application was coordinated and managed through this project. (author)

  14. Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.


    The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: (1) basic requirement for the integral reactor system design (2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system (3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs.

  15. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)


    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  16. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)


    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  17. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.


    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  18. The MAUS nuclear space reactor with ion propulsion system (United States)

    Mainardi, Enrico


    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome "La Sapienza" starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  19. Enhanced Combustion Low NOx Pulverized Coal Burner

    Energy Technology Data Exchange (ETDEWEB)

    David Towle; Richard Donais; Todd Hellewell; Robert Lewis; Robert Schrecengost


    For more than two decades, Alstom Power Inc. (Alstom) has developed a range of low cost, infurnace technologies for NOx emissions control for the domestic U.S. pulverized coal fired boiler market. This includes Alstom's internally developed TFS 2000{trademark} firing system, and various enhancements to it developed in concert with the U.S. Department of Energy. As of the date of this report, more than 270 units representing approximately 80,000 MWe of domestic coal fired capacity have been retrofit with Alstom low NOx technology. Best of class emissions range from 0.18 lb/MMBtu for bituminous coal to 0.10 lb/MMBtu for subbituminous coal, with typical levels at 0.24 lb/MMBtu and 0.13 lb/MMBtu, respectively. Despite these gains, NOx emissions limits in the U.S. continue to ratchet down for new and existing boiler equipment. On March 10, 2005, the Environmental Protection Agency (EPA) announced the Clean Air Interstate Rule (CAIR). CAIR requires 25 Eastern states to reduce NOx emissions from the power generation sector by 1.7 million tons in 2009 and 2.0 million tons by 2015. Low cost solutions to meet such regulations, and in particular those that can avoid the need for a costly selective catalytic reduction system (SCR), provide a strong incentive to continue to improve low NOx firing system technology to meet current and anticipated NOx control regulations. The overall objective of the work is to develop an enhanced combustion, low NOx pulverized coal burner, which, when integrated with Alstom's state-of-the-art, globally air staged low NOx firing systems will provide a means to achieve: Less than 0.15 lb/MMBtu NOx emissions when firing a high volatile Eastern or Western bituminous coal, Less than 0.10 lb/MMBtu NOx emissions when firing a subbituminous coal, NOx reduction costs at least 25% lower than the costs of an SCR, Validation of the NOx control technology developed through large (15 MWt) pilot scale demonstration, and Documentation required for

  20. Reactor/Brayton power systems for nuclear electric spacecraft (United States)

    Layton, J. P.


    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  1. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)


    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  2. Reliability of digital reactor protection system based on extenics. (United States)

    Zhao, Jing; He, Ya-Nan; Gu, Peng-Fei; Chen, Wei-Hua; Gao, Feng


    After the Fukushima nuclear accident, safety of nuclear power plants (NPPs) is widespread concerned. The reliability of reactor protection system (RPS) is directly related to the safety of NPPs, however, it is difficult to accurately evaluate the reliability of digital RPS. The method is based on estimating probability has some uncertainties, which can not reflect the reliability status of RPS dynamically and support the maintenance and troubleshooting. In this paper, the reliability quantitative analysis method based on extenics is proposed for the digital RPS (safety-critical), by which the relationship between the reliability and response time of RPS is constructed. The reliability of the RPS for CPR1000 NPP is modeled and analyzed by the proposed method as an example. The results show that the proposed method is capable to estimate the RPS reliability effectively and provide support to maintenance and troubleshooting of digital RPS system.

  3. N-reactor charge-discharge system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tokarz, R.D.; Marr, G.D.; Nesbitt, J.F.


    This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies.

  4. 30 CFR 56.7803 - Lighting the burner. (United States)


    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Lighting the burner. 56.7803 Section 56.7803... Piercing Rotary Jet Piercing § 56.7803 Lighting the burner. A suitable means of protection shall be provided for the employee when lighting the burner. ...

  5. 30 CFR 57.7803 - Lighting the burner. (United States)


    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Lighting the burner. 57.7803 Section 57.7803... Jet Piercing Rotary Jet Piercing-Surface Only § 57.7803 Lighting the burner. A suitable means of protection shall be provided for the employee when lighting the burner. ...

  6. Supervisory Control System Architecture for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL


    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  7. Compatibility of refractory materials for nuclear reactor poison control systems (United States)

    Sinclair, J. H.


    Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

  8. Nuclear plant-aging research on reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.C.


    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  9. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)


    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  10. Numerical simulation of the power characteristics of twin-core pulse reactor-pumped laser system (United States)

    Gulevich, A. V.; Barzilov, A. P.; Dyachenko, P. P.; Zrodnikov, A. V.; Kukharchuk, O. F.; Kachanov, B. V.; Kolyada, S. G.; Pashin, E. A.


    Concept for high-power pulsed reactor-pumped laser system (RPLS) based on the new physical principles (direct nuclear-to-optical conversion) is discussed with reference to ICF feasibility problem. Theoretical problems for substantiation of the neutronic and physical characteristics of the RPLS power model are considered. Results of numerical studies of the expected power characteristics of reactor laser system are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.


    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.

  12. Bed burners for grate boilers; Baeddbraennare foer rosterpannor

    Energy Technology Data Exchange (ETDEWEB)

    Sendelius, Mikael; Schuster, Robert [AaF-Energikonsult AB, Stockholm (Sweden)


    The objective of this work is to increase the knowledge of bed burners and their optimal positions in furnaces. The results from several computational fluid mechanics calculations are presented. An investigation concerning bed burners among plant owners is included as well. A bed burner is defined as a burner used for enhancing the combustion process on the bed i.e. it is used to dry incoming wet fuel. A load burner is used to quickly increase the boiler load and primarily not for creating better combustion conditions on the grate. Fluid mechanics calculations have been performed for five different cases, including the reference case. The following four bed burner arrangements have been examined: flat flame burner, six burners placed in the combustion chamber, two symmetric placed burners and two asymmetric placed burners. The same furnace model has been used through all the simulations. The incident radiation has been calculated in order to determine which one of the bed burners having the best possibility to improve the combustion process on the grate. The results showed that the flat flame burner and the six burners placed in the combustion chamber gave the most incident radiation on the first two grate zones. Bed burners placed further back in the furnace gave less good results. A comparison between the reference case (the case without burners) and the case with two burners showed that there was almost no difference in incident radiation between the two cases. The case with six burners placed in the combustion chamber gave most incident radiation, however this arrangement gave an irregular distribution of the radiation on the bed. Too high or irregular distributed radiation increases the risk for getting regions, on the grate, where the fuel is completely burnt. Primary air will pass through these regions. This phenomenon will lead to high temperatures that cause increased levels of emissions, in particular NO{sub x}. Reorganizing the burner positions and

  13. Pressure Melting and Ice Skating / Bunsen Burner

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 1; Issue 5. Pressure Melting and Ice Skating / Bunsen Burner - Revisited. Classroom Volume 1 Issue 5 May 1996 pp 71-78. Fulltext. Click here to view fulltext PDF. Permanent link: Resonance ...

  14. On Bunsen Burners, Bacteria and the Bible

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 1; Issue 2. On Bunsen Burners, Bacteria and the Bible. Milind Watve. Classroom Volume 1 Issue 2 February 1996 pp 84-89. Fulltext. Click here to view fulltext PDF. Permanent link: ...

  15. On feasibility of optimizing the neutronic parameters of a laser system pumped by a pulsed reactor


    A.V. Gulevich; O.F. Kukharchuk; A.I. Brezhnev; A.A. Suvorov


    The paper examines the calculated feasibility of improving the energy characteristics of power pulses in a system consisting of a reactor and a subcritical block. A BARS-type fast neutron reactor is used as a self-quenching pulsed reactor. The subcritical block is a cylindrical structure comprising laser-active elements, moderator components and two reflectors (internal and external). The internal reflector material is zirconium hydride, and the external reflector material is beryllium. Th...

  16. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A


    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  17. Development of an advanced high efficiency coal combustor for boiler retrofit. Task 1, Cold flow burner development: Final report

    Energy Technology Data Exchange (ETDEWEB)

    LaFlesh, R.C.; Rini, M.J.; McGowan, J.G.


    The overall objective of this program is to develop a high efficiency advanced coal combustor (HEACC) for coal-based fuels capable of being retrofitted to industrial boilers originally designed for firing natural gas, distillate, and/or residual oil. The HEACC system is to be capable of firing microfine coal water fuel (MCWF), MCWF with alkali sorbent (for SO{sub 2} reduction), and dry microfine coal. Design priorities for the system are that it be simple to operate and will offer significant reductions in NO{sub x}, SO{sub x}, and particulate emissions as compared with current coal fired combustor technology. The specific objective of this report is to document the work carried out under Task 1.0 of this contract, ``Cold Flow Burner Development``. As are detailed in the report, key elements of this work included primary air swirler development, burner register geometry design, cold flow burner model testing, and development of burner scale up criteria.

  18. Reactor safeguards

    CERN Document Server

    Russell, Charles R


    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  19. Market assessment for the fan atomized oil burner

    Energy Technology Data Exchange (ETDEWEB)

    Westphalen, D. [A.D. Little, Inc., Cambridge, MA (United States)


    The market potential for the fan atomized burner (FAB) in water and space heating applications was examined. The major findings of the study are as follows. (1). The FAB`s low-input capability allows development of oil-fired room heaters and wall furnaces, a new market area for oil heat. (2). Among conventional oil-fired products, furnaces will benefit most from the burner`s low input capability due to (1) their quick delivery of heat and (2) their more prevalent use in warmer climates and smaller homes. (3). The greatest potential for increased product sales or oil sales exists in the use of the burner with new products (i.e., room heaters). Sales of boilers and direct-fired water heaters are not likely to increase with the use of the burner. (4). Acceptance of the burner will be dependent on proof of reliability. Proof of better reliability than conventional burners would accelerate acceptance.

  20. Shielding considerations for advanced space nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, J.P. Jr.; Buden, D.


    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  1. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor (United States)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed


    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.


    Directory of Open Access Journals (Sweden)



    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  3. The fluidized bed reactor with a prepolymerization system and its influence on polymer physicochemical characteristics

    Directory of Open Access Journals (Sweden)

    Fernandes F.A.N.


    Full Text Available This work addresses the influence of a prepolymerization system on the behavior of the fluidized bed reactor used for polyethylene production. Its influence on the polymer's physicochemical characteristics and production was also studied. The results indicate that the use of prepolymerized catalyst particles results in milder temperatures in the fluidized bed reactor, thus avoiding the formation of hot spots, melting of the polymer particle and reactor shutdown. Productivity can be enhanced depending on the operational conditions used in the prepolymerization reactor.

  4. Increased Coal Replacement in a Cement Kiln Burner by Feeding a Mixture of Solid Hazardous Waste and Shredded Plastic Waste


    Ariyaratne, W.K. Hiromi; Melaaen, Morten Christian; Tokheim, Lars-André


    The present study aims to find the maximum possible replacement of coal by combined feeding of plastic waste and solid hazardous waste mixed with wood chips (SHW) in rotary kiln burners used in cement kiln systems. The coal replacement should be achieved without negative impacts on product quality, emissions or overall operation of the process. A full-scale experiment was carried out in the rotary kiln burner of a cement kiln by varying SHW and plastic waste feeding rates. Experimental result...

  5. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin


    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  6. BEACON TSM application system to the operation of PWR reactors; Aplicacion del Sistema BEACON TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.


    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  7. Pressurized hydrogenotrophic denitrification reactor for small water systems. (United States)

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal


    The implementation of hydrogenotrophic denitrification is limited due to safety concerns, poor H2 utilization and low solubility of H2 gas with the resulting low transfer rate. The current paper presents the main research work conducted on a pressurized hydrogenotrophic reactor for denitrification that was recently developed. The reactor is based on a new concept suggesting that a gas-liquid equilibrium is achieved in the closed headspace of denitrifying reactor, further produced N2 gas is carried out by the effluent and gas purging is not required. The feasibility of the proposed reactor was shown for two effluent concentrations of 10 and 1 mg NO3--N/L. Hydrogen gas utilization efficiencies of 92.8% and 96.9% were measured for the two effluent concentrations, respectively. Reactor modeling predicted high denitrification rates above 4 g NO3--N/(Lreactor·d) at reasonable operational conditions. Hydrogen utilization efficiency was improved up to almost 100% by combining the pressurized reactor with a following open-to-atmosphere polishing unit. Also, the potential of the reactor to remove ClO4- was shown. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. Measurements of the concentration of major chemical species in the flame of a test burner with a air swirling system; Mesures de concentration d`especes chimiques majoritaires dans la flamme d`un bruleur modele avec mise en rotation de l`air

    Energy Technology Data Exchange (ETDEWEB)

    Albert, St. [Gaz de France (GDF), 93 - La Plaine-Saint-Denis (France); Most, J.M.; Poireault, B. [Centre National de la Recherche Scientifique (CNRS), 86 - Poitiers (France)


    The study of combustion in industrial burners remains difficult because of the complexity of the equipments used: materials geometry, tri-dimensional flows etc.. The phenomena that control the combustion in a gas burner with a swirl air system has been studied thanks to a collaboration between the Direction of Research of Gaz de France (GdF) and the Laboratory for Combustion and Detonation Research (LCD) of the French National Centre of Scientific Research (CNRS). The burner used is developed by the LCD and the measurements of stable chemical species were performed by the CERSTA centre of GdF. These series of tests, performed in confined environment, have permitted to identify some of the parameters that influence combustion chemistry. Mapping of chemical species allows to distinguish 5 zones of flame development and also the zones of nitrogen oxides formation. Methane is rapidly centrifuged a few millimeters above the injection pipe and centrifuged with rotating combustion air. Carbon monoxide occurs immediately in the central recirculation zone which is weakly reactive (no oxygen and no methane). Oxygen content increases downflow from this area and carbon dioxide reaches its concentration maxima. CO formation decreases when the swirl number increases and CO{sub 2} formation occurs earlier. On the contrary, the emissions of CO and CH{sub 4} do not depend on the swirl value and the NO{sub x} values are only slightly dependent on this value. (J.S.)

  9. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira


    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  10. Development of lean premixed low-swirl burner for low NO{sub x} practical application

    Energy Technology Data Exchange (ETDEWEB)

    Yegian, D.T.; Cheng, R.K.


    Laboratory experiments have been performed to evaluate the performance of a premixed low-swirl burner (LSB) in configurations that simulate commercial heating appliances. Laser diagnostics were used to investigate changes in flame stabilization mechanism, flowfield, and flame stability when the LSB flame was confined within quartz cylinders of various diameters and end constrictions. The LSB adapted well to enclosures without generating flame oscillations and the stabilization mechanism remained unchanged. The feasibility of using the LSB as a low NO{sub x} commercial burner has also been verified in a laboratory test station that simulates the operation of a water heater. It was determined that the LSB can generate NO{sub x} emissions < 10 ppm (at 3% O{sub 2}) without significant effect on the thermal efficiency of the conventional system. The study has demonstrated that the lean premixed LSB has commercial potential for use as a simple economical and versatile burner for many low emission gas appliances.

  11. Performance test reports and comparison of emission characteristics of prototype liquid multifuel burners developed for US military field cooking applications

    Energy Technology Data Exchange (ETDEWEB)

    Litzke, W.; Celebi, Y.; McDonald, R.


    The objective of this project is to provide data to the U.S. Army Natick RD&E Center on the performance of three prototype burners, which have the capability of firing with multiple types of fuels (diesel and JP-8), and the conventional gasoline-fired M-2 burner. The prototype burners are intended to replace the M-2 unit currently used in food cooking appliances in the Army. The burners supplied to Brookhaven National Laboratory (BNL) for the purpose of testing under this project included one M-2 unit, one M-3 prototype unit designed by Natick, one Babington prototype unit designed by Babington Engineering, and one ITR prototype designed by International Thermal Research Ltd. It should be noted, however, that after the project began, Babington Engineering provided an upgraded prototype unit for testing which replaced the unit initially provided by the Natick Center. The M-3 unit replaced the Karcher unit listed in the contract. The test procedures which were described in a Test Method Report allowed for the measurement of the concentrations of specific compounds emitted from the burners. These compounds included oxygen (O{sub 2}), carbon monoxide (CO), oxides of nitrogen (NOx), formaldehyde, and particulate emissions. The level of smoke produced was also measured by using a Bacharach Smoke Number system (ASTM Standard D2156). A separate Performance Test Report for each burner was prepared as part of this project, and is attached as part of this report. In those reports details of the measurement techniques, instrumentation, test operating conditions, and data for each burner were included. This paper provides a summary and a comparison of the results for all burners. A brief discussion of emissions from other similar small oil combustion systems is also part of this document to provide perspective on the type of contaminants and levels expected from these systems.

  12. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  13. A small, 1400 deg Kelvin, reactor for Brayton space power systems (United States)

    Lantz, E.; Mayo, W.


    A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

  14. A pragmatic approach towards designing a second shutdown system for Tehran research reactor


    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein


    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in man...

  15. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors (United States)


    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.'' DG-1277...

  16. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (United States)


    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors...

  17. Fuel burner and combustor assembly for a gas turbine engine (United States)

    Leto, Anthony


    A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

  18. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor (United States)

    Rohanda, Anis; Waris, Abdul


    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  19. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system (United States)

    Jefferies, K. S.; Tew, R. C.


    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  20. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)


    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.


    Energy Technology Data Exchange (ETDEWEB)

    Robert States


    Non steady impingement heat transfer is measured. Impingement heating consumes 130 T-BTU/Yr in paper drying, but is only 25% thermally efficient. Pulse impingement is experimentally shown to enhance heat transfer by 2.8, and may deliver thermal efficiencies near 85%. Experimental results uncovered heat transfer deviations from steady theory and from previous investigators, indicating the need for further study and a better theoretical framework. The pulse burner is described, and its roll in pulse impingement is analyzed.

  2. Investigation of Anaerobic Fluidized Bed Reactor/ Aerobic Moving Bed Bio Reactor (AFBR/MMBR) System for Treatment of Currant Wastewater (United States)

    JAFARI, Jalil; MESDAGHINIA, Alireza; NABIZADEH, Ramin; FARROKHI, Mehrdad; MAHVI, Amir Hossein


    Background: Anaerobic treatment methods are more suitable for the treatment of concentrated wastewater streams, offer lower operating costs, the production of usable biogas product. The aim of this study was to investigate the performance of an Anaerobic Fluidized Bed Reactor (AFBR)-Aerobic Moving Bed Bio Reactor (MBBR) in series arrangement to treat Currant wastewater. Methods: The bed materials of AFBR were cylindrical particles made of PVC with a diameter of 2–2.3 mm, particle density of 1250 kg/m3. The volume of all bed materials was 1.7 liter which expanded to 2.46 liters in fluidized situation. In MBBR, support media was composed of 1.5 liters Bee-Cell 2000 having porosity of 87% and specific surface area of 650m2/m3. Results: When system operated at 35 ºC, chemical oxygen demand (COD) removal efficiencies were achieved to 98% and 81.6% for organic loading rates (OLR) of 9.4 and 24.2 g COD/l.d, and hydraulic retention times (HRT) of 48 and 18 h, in average COD concentration feeding of 18.4 g/l, respectively. Conclusion: The contribution of AFBR in total COD removal efficiency at an organic loading rate (OLR) of 9.4 g COD/l.d was 95%, and gradually decreased to 76.5% in OLR of 24.2 g COD/l.d. Also with increasing in organic loading rate the contribution of aerobic reactor in removing COD gradually decreased. In this system, the anaerobic reactor played the most important role in the removal of COD, and the aerobic MBBR was actually needed to polish the anaerobic treated wastewater. PMID:26056640


    Energy Technology Data Exchange (ETDEWEB)

    Hamid Farzan; Jennifer Sivy; Alan Sayre; John Boyle


    Under sponsorship of the Department of Energy's National Energy Technology Laboratory (NETL), McDermott Technology, Inc. (MTI), the Babcock & Wilcox Company (B&W), and Fuel Tech teamed together to investigate an integrated solution for NOx control. The system was comprised of B&W's DRB-4Z{trademark} low-NO{sub x} pulverized coal (PC) burner technology and Fuel Tech's NO{sub x}OUT{reg_sign}, a urea-based selective non-catalytic reduction (SNCR) technology. The technology's emission target is achieving 0.15 lb NO{sub x}/10{sup 6} Btu for full-scale boilers. Development of the low-NOx burner technology has been a focus in B&W's combustion program. The DRB-4Z{trademark} burner (see Figure 1.1) is B&W's newest low-NO{sub x} burner capable of achieving very low NO{sub x}. The burner is designed to reduce NO{sub x} by diverting air away from the core of the flame, which reduces local stoichiometry during coal devolatilization and, thereby, reduces initial NO{sub x} formation. Figure 1.2 shows the historical NO{sub x} emission levels from different B&W burners. Figure 1.2 shows that based on three large-scale commercial installations of the DRB-4Z{trademark} burners in combination with OFA ports, using Western subbituminous coal, the NO{sub x} emissions ranged from 0.16 to 0.18 lb/10{sup 6} Btu. It appears that with continuing research and development the Ozone Transport Rule (OTR) emission level of 0.15 lb NO{sub x}/10{sup 6} Btu is within the reach of combustion modification techniques for boilers using western U.S. subbituminous coals. Although NO{sub x} emissions from the DRB-4Z{trademark} burner are nearing OTR emission level with subbituminous coals, the utility boiler owners that use bituminous coals can still benefit from the addition of an SNCR and/or SCR system in order to comply with the stringent NO{sub x} emission levels facing them.

  4. Coal-water mixture fuel burner (United States)

    Brown, T.D.; Reehl, D.P.; Walbert, G.F.


    The present invention represents an improvement over the prior art by providing a rotating cup burner arrangement for use with a coal-water mixture fuel which applies a thin, uniform sheet of fuel onto the inner surface of the rotating cup, inhibits the collection of unburned fuel on the inner surface of the cup, reduces the slurry to a collection of fine particles upon discharge from the rotating cup, and further atomizes the fuel as it enters the combustion chamber by subjecting it to the high shear force of a high velocity air flow. Accordingly, it is an object of the present invention to provide for improved combustion of a coal-water mixture fuel. It is another object of the present invention to provide an arrangement for introducing a coal-water mixture fuel into a combustion chamber in a manner which provides improved flame control and stability, more efficient combustion of the hydrocarbon fuel, and continuous, reliable burner operation. Yet another object of the present invention is to provide for the continuous, sustained combustion of a coal-water mixture fuel without the need for a secondary combustion source such as natural gas or a liquid hydrocarbon fuel. Still another object of the present invention is to provide a burner arrangement capable of accommodating a coal-water mixture fuel having a wide range of rheological and combustion characteristics in providing for its efficient combustion. 7 figs.

  5. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila


    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  6. Application of neutron activation analysis system in Xi'an pulsed reactor

    CERN Document Server

    Zhang Wen Shou; Yu Qi


    Neutron Activation Analysis System in Xi'an Pulsed Reactor is consist of rabbit fast radiation system and experiment measurement system. The functions of neutron activation analysis are introduced. Based on the radiation system. A set of automatic data handling and experiment simulating system are built. The reliability of data handling and experiment simulating system had been verified by experiment

  7. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor (United States)

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.


    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  8. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)


    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

  9. Modelling of Mass Transfer Phenomena in Chemical and Biochemical Reactor Systems using Computational Fluid Dynamics

    DEFF Research Database (Denmark)

    Larsson, Hilde Kristina

    the velocity and pressure distributions in a fluid. CFD also enables the modelling of several fluids simultaneously, e.g. gas bubbles in a liquid, as well as the presence of turbulence and dissolved chemicals in a fluid, and many other phenomena. This makes CFD an appreciated tool for studying flow structures......, mixing, and other mass transfer phenomena in chemical and biochemical reactor systems. In this project, four selected case studies are investigated in order to explore the capabilities of CFD. The selected cases are a 1 ml stirred microbioreactor, an 8 ml magnetically stirred reactor, a Rushton impeller...... stirred pilot plant reactor, and a rotating bed reactor filled with catalytic porous material. A selection of the simulated phenomena includes the velocities and turbulent quantities in the reactors, as well as the distribution of the gas and liquid phases in them. Mixing times, oxygen transfer rates...

  10. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others


    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  11. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Suzuki, Katsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  12. System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.H.; Lewis, B.R.; Klein, A.C. (Department of Nuclear Engineering, Oregon State University, Radiation Center, C116, Corvallis, Oregon 97331-5902 (United States)); Pawlowski, R.A. (Battelle Pacific Northwest Laboratories, Richland, Washington 99352 (United States))


    Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range.

  13. High Efficiency Microchannel Sabatier Reactor System for In Situ Resource Utilization Project (United States)

    National Aeronautics and Space Administration — An innovative Microchannel Sabatier Reactor System (MSRS) is proposed for 100% recovery of oxygen (as water) and methane from carbon dioxide (CO2), a valuable in...

  14. Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim


    Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  15. Modeling, simulation, and optimization of a front-end system for acetylene hydrogenation reactors

    Directory of Open Access Journals (Sweden)

    R. Gobbo


    Full Text Available The modeling, simulation, and dynamic optimization of an industrial reaction system for acetylene hydrogenation are discussed in the present work. The process consists of three adiabatic fixed-bed reactors, in series, with interstage cooling. These reactors are located after the compression and the caustic scrubbing sections of an ethylene plant, characterizing a front-end system; in contrast to the tail-end system where the reactors are placed after the de-ethanizer unit. The acetylene conversion and selectivity profiles for the reactors are optimized, taking into account catalyst deactivation and process constraints. A dynamic optimal temperature profile that maximizes ethylene production and meets product specifications is obtained by controlling the feed and intercoolers temperatures. An industrial acetylene hydrogenation system is used to provide the necessary data to adjust kinetics and transport parameters and to validate the approach.

  16. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    National Research Council Canada - National Science Library

    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein


    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall...

  17. Progress in space nuclear reactor power systems technology development - The SP-100 program (United States)

    Davis, H. S.


    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  19. The detector system of the Daya Bay reactor neutrino experiment


    An, F. P.; Carr, R.; McKeown, R.D.; Tsang, R. H. M.; Wu, F.F.


    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world׳s most precise measurement of sin^2 2θ_(13) and the effective mass splitting Δm^2_(ee). The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world׳s most prolific sources of electron antineutrinos. Multiple antineutrino detect...

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

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz


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

  1. Perbandingan Unjuk Kerja Kompor Methanol dengan Variasi Diameter Burner

    Directory of Open Access Journals (Sweden)

    Subroto Subroto


    Full Text Available Sampai saat ini sebagian masyarakat masih anyak menggunakan minyak tanah sebagai bahan bakar untuk keperluan rumah tangga maupun industri, walaupun harga minyak tanah naik cukup tinggi karena subsidi dari pemerintah dicabut. Untuk mengurangi ketergantungan minyak tanah perlu penggunaan bahan bakar alternatif yaitu methanol. Methanol mempunyai kelebihan mudah didapatkan di lapangan dan dengan harga yang lebih murah dari minyak tanah. Kompor methanol sudah dikenal masyarakat akan tetapi penggunaannya masih sangat terbatas karena unjuk kerjanya masih kurang baik dibandingkan kompor minyak tanah. Penelitian ini bertujuan untuk mengetahui unjuk kerja kompor methanol melalui pengaruh variasi diameter burner.               Penelitian dimulai dengan rancang bangun burner terbuat dari bahan kuningan dengan tiga macam model dengan variasi diameter burner dan tinggi burner maupun jumlah lubang tetap. Pengujian unjuk kerja berdasarkan karakteristik pembakaran dilakukan melalui water boiling test. Parameter yang diukur meliputi temperatur api pembakaran, temperatur air, konsumsi bahan bakar, dan waktu pendidihan.               Hasil penelitian menunjukkan bahwa variasi diameter burner berpengaruh terhadap karakteristik pembakaran yang dihasilkan. Temperatur pembakaran yang tinggi dicapai oleh burner dengan diameter 12,8mm dan 10mm, konsumsi bahan bakar yang kecil burner diameter 12,8mm dan waktu pendidihan yang pendek dicapai burner 12,8mm. Jadi kompor methanol dengan unjuk kerja terbaik adalah kompor methanol dengan diameter burner 12,8mm.

  2. Metal fire implications for advanced reactors. Part 1, literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.


    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  3. Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Son, Han Seong; Song, Deok Yong [ENESYS, Daejeon (Korea, Republic of)


    As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code.

  4. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor


    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  5. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)


    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  6. Research and Development of Natural Draft Ultra-Low Emissions Burners for Gas Appliances

    Energy Technology Data Exchange (ETDEWEB)

    Therkelsen, Peter [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Cheng, Robert [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sholes, Darren [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)


    Combustion systems used in residential and commercial cooking appliances must be robust and easy to use while meeting air quality standards. Current air quality standards for cooking appliances are far greater than other stationary combustion equipment. By developing an advanced low emission combustion system for cooking appliances, the air quality impacts from these devices can be reduced. This project adapted the Lawrence Berkeley National Laboratory (LBNL) Ring-Stabilizer Burner combustion technology for residential and commercial natural gas fired cooking appliances (such as ovens, ranges, and cooktops). LBNL originally developed the Ring-Stabilizer Burner for a NASA funded microgravity experiment. This natural draft combustion technology reduces NOx emissions significantly below current SCAQMD emissions standards without post combustion treatment. Additionally, the Ring-Stabilizer Burner technology does not require the assistance of a blower to achieve an ultra-low emission lean premix flame. The research team evaluated the Ring-Stabilizer Burner and fabricated the most promising designs based on their emissions and turndown.

  7. High strength sewage treatment in a UASB reactor and an integrated UASB-digester system. (United States)

    Mahmoud, Nidal


    The treatment of high strength sewage was investigated in a one-stage upflow anaerobic sludge blanket (UASB) reactor and a UASB-digester system. The one-stage UASB reactor was operated in Palestine at a hydraulic retention time (HRT) of 10h and at ambient air temperature for a period of more than a year in order to asses the system response to the Mediterranean climatic seasonal temperature fluctuation. Afterwards, the one-stage UASB reactor was modified to a UASB-digester system by incorporating a digester operated at 35 degrees C. The achieved removal efficiencies in the one-stage UASB reactor for total, suspended, colloidal, dissolved and VFA COD were 54, 71, 34, 23%, and -7%, respectively during the first warm six months of the year, and achieved only 32% removal efficiency for COD total over the following cold six months of the year. The modification of the one-stage UASB reactor to a UASB-digester system had remarkably improved the UASB reactor performance as the UASB-digester achieved removal efficiencies for total, suspended, colloidal, dissolved and VFA COD of 72, 74, 74, 62 and 70%. Therefore, the anaerobic treatment of high strength sewage during the hot period in Palestine in a UASB-digester system is very promising.

  8. Method and apparatus for enhancing reactor air-cooling system performance (United States)

    Hunsbedt, A.


    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  9. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (United States)


    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  10. Burners for Supersonic Ramjets - Some Observations on Instability in a Two-Inch Ramjet Burner (United States)


    a burner comprising a simple 0.75" oxyhydrogen pilot cone (Igniter 1-2, Figure 3) is much more prone to insta- bility than one comprising the same...occurred largely with a "poor igniter." Most of the work described above was done with a simple 0.75 inch oxyhydrogen cone. 2. It was accompanied

  11. System and method for producing substitute natural gas from coal (United States)

    Hobbs, Raymond [Avondale, AZ


    The present invention provides a system and method for producing substitute natural gas and electricity, while mitigating production of any greenhouse gasses. The system includes a hydrogasification reactor, to form a gas stream including natural gas and a char stream, and an oxygen burner to combust the char material to form carbon oxides. The system also includes an algae farm to convert the carbon oxides to hydrocarbon material and oxygen.

  12. Fossil-fuel processing technical/professional services: comparison of Fischer-Tropsch reactor systems. Phase I, final report

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, G.J.; Riekena, M.L.; Vickers, A.G.


    The Fischer-Tropsch reaction was commercialized in Germany and used to produce military fuels in fixed bed reactors. It was recognized from the start that this reactor system had severe operating and yield limitations and alternative reactor systems were sought. In 1955 the Sasol I complex, using an entrained bed (Synthol) reactor system, was started up in South Africa. Although this reactor was a definite improvement and is still operating, the literature is filled with proponents of other reactor systems, each claiming its own advantages. This report provides a summary of the results of a study to compare the development potential of three of these reactor systems with the commercially operating Synthol-entrained bed reactor system. The commercial Synthol reactor is used as a benchmark against which the development potential of the other three reactors can be compared. Most of the information on which this study is based was supplied by the M.W. Kellogg Co. No information beyond that in the literature on the operation of the Synthol reactor system was available for consideration in preparing this study, nor were any details of the changes made to the original Synthol system to overcome the operating problems reported in the literature. Because of conflicting claims and results found in the literature, it was decided to concentrate a large part of this study on a kinetic analysis of the reactor systems, in order to provide a theoretical analysis of intrinsic strengths and weaknesses of the reactors unclouded by different catalysts, operating conditions and feed compositions. The remainder of the study considers the physical attributes of the four reactor systems and compares their respective investment costs, yields, catalyst requirements and thermal efficiencies from simplified conceptual designs.

  13. A system dynamics model for tritium cycle of pulsed fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)


    As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

  14. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun


    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

  15. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.


    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  16. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  17. Challenges to deployment of twenty-first century nuclear reactor systems. (United States)

    Ion, Sue


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

  18. Challenges to deployment of twenty-first century nuclear reactor systems (United States)

    Ion, Sue


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

  19. Challenges to deployment of twenty-first century nuclear reactor systems (United States)


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors. PMID:28293142

  20. Control of flames by tangential jet actuators in oxy-fuel burners

    Energy Technology Data Exchange (ETDEWEB)

    Boushaki, Toufik [CORIA UMR 6614 CNRS-Universite et INSA de ROUEN, Site Universitaire du Madrillet, 76801 Saint Etienne du Rouvray, Cedex (France); Universite de Toulouse-INPT-UPS, IMFT (Institut de Mecanique des Fluides de Toulouse), Allee Camille Soula, F-31400 Toulouse, Cedex (France); Sautet, Jean-Charles [CORIA UMR 6614 CNRS-Universite et INSA de ROUEN, Site Universitaire du Madrillet, 76801 Saint Etienne du Rouvray, Cedex (France); Labegorre, Bernard [Air Liquide, Centre de Recherche Claude-Delorme, Les Loges-en-Josas, B.P. 126 78354 Jouy-en-Josas, Cedex (France)


    The active control of oxy-fuel flames from burners with separated jets is investigated. The control system consists of four small jet actuators, placed tangential to the exit of the main jets to generate a swirling flow. These actuators are able to modify the flow structure and to act on mixing between the reactants and consequently on the flame behavior. The burner (25 kW) is composed of separated jets, one jet of natural gas and one or two jets of pure oxygen. Experiments are conducted with three burner configurations, according to the number of jets, the jet exit velocities, and the separation distance between the jets. OH chemiluminescence measurements, particle image velocimetry, and measurements of NO{sub x} emissions are used to characterize the flow and the flame structure. Results show that the small jet actuators have a significant influence on the behavior of jets and the flame characteristics, particularly in the stabilization zone. It is shown that the control leads to a decrease in lift-off heights and to better stability of the flame. The use of jet actuators induces high jet spreading and an increase in turbulence intensity, which improves the mixing between the reactants and the surrounding fluid. Pollutant measurements show important results in terms of NO{sub x} reductions (up to 60%), in particular for low swirl intensity. The burner parameters, such as the number of jets and the spacing between the jets, also impact the flame behavior and NO{sub x} formation. (author)

  1. Some parameters and conditions defining the efficiency of burners ...

    Indian Academy of Sciences (India)

    A number of new wordings and statements regarding the targeted problem of destruction of long-lived wastes (transmutation) is considered. Some new criteria concerning the efficiency of a particular burner type are proposed. It is shown that the destruction efficiency of a specific burner is greatly influenced by the ...

  2. Numerical simulation of Venturi ejector reactor in yellow phosphorus purification system

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiao-jing; Tang, Lei, E-mail:; Jiang, Zeng


    Highlights: • Venturi ejector reactor is used in yellow phosphorus purification system to obtain high purity phosphorus. • We study the changes of vacuum region and the performances of Venturi ejector reactor with different operating pressure. • The whole study is aim to investigate the operating conditions, rather than to find out the small details of the chemical reaction. - Abstract: A novel type of Venturi ejector reactor, which was used in a pilot plant test in a factory in Guizhou in China, was developed to overcome the insufficiency of chemical reaction in the stirred-tank reactor in yellow phosphorus purification system. The effects of different working medium, the changes of vacuum region, and the performances of the Venturi ejector reactor with different operating pressure were investigated by FLUENT. Results show that the absolute value of vacuum pressure of single-phase flow was smaller than two-phase flow at the same operating conditions, which meat two-phase flow has a higher suction capability. Reflow phenomena occurred near the exit of suction pipe and nozzle. The former reflow which leads to energy loss of vacuum region was undesirable, and the latter was beneficial to the dispersion of liquid yellow phosphorus. With a flow rate ratio below 0.45, the performance of the Venturi ejector reactor was effective. By adjusting the operating pressure, a proper flow rate ratio could be satisfied to meet the production needs in yellow phosphorus purification system.

  3. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)


    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  4. Graphic-object information system {open_quotes}research base for reactor materials science{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Markina, N.V.; Lebedeva, E.E.; Arkhangel`skii, N.V.; Semenov, S.B.; Moiseev, A.L.


    An information system developed for reactor materials research is described. The information system incorporates an expert system, MATREKS, and a heirarchial data base. The data base contains information from 20 Russian research reactors. The information system structure, data base structure, search methods, system output modes, and technical facilities and software required are briefly discussed. 6 refs., 2 figs.

  5. Reference reactor module for NASA's lunar surface fission power system

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL


    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  6. Analysis of space reactor system components: Investigation through simulation and non-nuclear testing (United States)

    Bragg-Sitton, Shannon M.

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

  7. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion (United States)

    George, Jeffrey A.


    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  8. Selection of power plant elements for future reactor space electric power systems

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Bennett, G.A.; Copper, K.


    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

  9. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.


    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  10. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon


    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  11. Combustion starter and maintenance burner for pulverized solid fossil fuels, and combustion chamber equipped with burners of the kind

    Energy Technology Data Exchange (ETDEWEB)

    Vidal, J.; Malaubier, F.; Mevel, J-C.


    A power station boiler is described, which is fired by pulverized coal. The boiler has an axial conduit for feeding the pulverized fuel in a primary combustion supporting gas. The burner incorporates a flame detector and an igniter in the refractory tap-hole or in the axial conduit. The advantage of the burner is improved efficiency and reduced operating cost.

  12. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  13. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  14. Biodegradation of 2,4-Dinitrotoluene and 2,6-Dinitrotoluene in a Pilot-Scale Aerobic Slurry Reactor System (United States)


    consideration ............................................................................. 7 4 3.8.1 Design basis for a hypothetical full-scale bioslurry ...treatment system ......... 74 3.8.2 Cost elements associated with bioslurry reactor systems ............................. 78 3.8.3 Comparison with...hypothetical full-scale bioslurry treatment system The treatment cost of a slurry reactor system depends mainly on three process parameters: (1) solids

  15. Neutronics design study on a minor actinide burner for transmuting spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok


    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors. The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the doppler coefficient, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics. (author). 34 refs., 22 tabs., 14 figs.

  16. Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation (United States)


    Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individual component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.

  17. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  18. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud


    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  19. Nuclear reactor system study for NASA/JPL (United States)

    Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.


    Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

  20. The development of ex-core neutron flux monitoring system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S


    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

  1. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sayyah, A. [Department of Radiation Application, Shahid Beheshti University (Iran, Islamic Republic of); Rahmani, F., E-mail: [K.N. Toosi University of Technology, Department of Physics (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute (NSTRI) (Iran, Islamic Republic of)


    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  2. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.


    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  3. Premixed burner experiments: Geometry, mixing, and flame structure issues

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, A.K.; Lewis, M.J.; Gupta, M. [Univ of Maryland, College Park, MD (United States)] [and others


    This research program is exploring techniques for improved fuel-air mixing, with the aim of achieving combustor operations up to stoichiometric conditions with minimal NO x and maximum efficiency. The experimental studies involve the use of a double-concentric natural gas burner that is operable in either premixed or non-premixed modes, and the system allows systematic variation of equivalence ratio, swirl strength shear length region and flow momentum in each annulus. Flame structures formed with various combinations of swirl strengths, flow throughput and equivalence ratios in premixed mode show the significant impact of swirl flow distribution on flame structure emanating from the mixedness. This impact on flame structure is expected to have a pronounced effect on the heat release rate and the emission of NO{sub x}. Thus, swirler design and configuration remains a key factor in the quest for completely optimized combustion. Parallel numerical studies of the flow and combustion phenomena were carried out, using the RSM and thek-{epsilon} turbulence models. These results have not only indicated the strengths and limitations of CFD in performance and pollutants emission predictions, but have provided guidelines on the size and strength of the recirculation produced and the spatio-temporal structure of the combustion flowfield. The first stage of parametric studies on geometry and operational parameters at Morgan State University have culminated in the completion of a one-dimensional flow code that is integrated with a solid, virtual model of the existing premixed burner. This coupling will provide the unique opportunity to study the impact of geometry on the flowfield and vice-versa, with particular emphasis on concurrent design optimization.

  4. Fan Atomized Burner design advances & commercial development progress

    Energy Technology Data Exchange (ETDEWEB)

    Kamath, B. [Heat-Wise, Inc., Ridge, NY (United States); Butcher, T.A. [Brookhaven National Lab., Upton, NY (United States)


    As a part of the Oil Heat Research and Development program, sponsored by the US Department of Energy, Brookhaven National Laboratory (BNL) has an on-going interest in advanced combustion technologies. This interest is aimed at: improving the initial efficiency of heating equipment, reducing long term fouling and efficiency degradation, reducing air pollutant emissions, and providing practical low-firing rate technologies which may lead to new, high efficiency oil-fired appliances. The Fan-Atomized Burner (FAB) technology is being developed at BNL as part of this general goal. The Fan-Atomized Burner uses a low pressure, air atomizing nozzle in place of the high pressure nozzle used in conventional burners. Because it is air-atomized the burner can operate at low firing rates without the small passages and reliability concerns of low input pressure nozzles. Because it uses a low pressure nozzle the burner can use a fan in place of the small compressor used in other air-atomized burner designs. High initial efficiency of heating equipment is achieved because the burner can operate at very low excess air levels. These low excess air levels also reduce the formation of sulfuric acid in flames. Sulfuric acid is responsible for scaling and fouling of heat exchanger surfaces.

  5. An add-on system including a micro-reactor for an atr-ir spectrometer

    DEFF Research Database (Denmark)


    The invention relates to an add-on system for an attenuated total reflectance infrared (ATR-IR) spectrometer, the add-on system allowing for time-resolved in situ IR measurements of heterogeneous mixtures. The add-on device comprises a micro-reactor (300A) forming a sample cavity (305) when...

  6. Utilizing a Russian space nuclear reactor for a US space mission: Systems integration issues (United States)

    Reynolds, E.; Schaefer, E.; Polansky, G.; Lacy, J.; Bocharov, A.


    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft. One component is the Topaz 2 Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz 2 reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz 2 was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz 2, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons learned regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  7. Utilizing a Russian Space Nuclear Reactor for a United States Space Mission: Systems Integration Issues (United States)

    Reynolds, Edward; Schaefer, Edward; Polansky, Gary; Lacy, Jeff; Bocharov, Anatoly


    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a U.S. spacecraft. One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons learned regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  8. Design Improvement for the Reactor Trip Switchgear System for APR1400 Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Gyu; Choi, Woong Seock; Sohn, Se Do [KEPCO Engineering and Construction Co., Deajeon (Korea, Republic of)


    The Reactor Trip Switchgear System (RTSS) performs the function to open the Reactor Trip Circuit Breaker (RTCB) when the RTSS receives trip signals from the Plant Protection System (PPS). The RTSS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) receives the reactor trip signals from four independent PPS divisions and performs the function to interrupt power from the Motor Generator Set (MG Set) to the Digital Rod Control System (DRCS). The RTSS for SHN 1 and 2 consists of four Reactor Trip Switchgears (RTSGs) which form the selective 2-out-of-4 logic. The RTSS design for APR 1400 DC has been changed from selective 2-out-of-4 to full 2-out-of-4 logic by configuring two independent sets of RTSS for diversity. The RTSS with the full 2-out-of-4 logic decreases the chances of generating an inadvertent reactor trip by a failure during maintenance or testing. We expect this design change to contribute to enhancing the plant availability. After all, the quantitative reliability analysis will be necessary to visualize the degree of the plant availability enhancement from the design change described in this paper.

  9. Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

    Energy Technology Data Exchange (ETDEWEB)

    Pablo Rubiolo, Principal Investigator


    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  10. Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

    Directory of Open Access Journals (Sweden)

    Ricardo Andreola


    Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

  11. Lunar in-core thermionic nuclear reactor power system conceptual design (United States)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.


    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  12. Flashback Avoidance in Swirling Flow Burners

    Directory of Open Access Journals (Sweden)

    Vigueras-Zúñiga Marco Osvaldo


    Full Text Available Lean premixed combustion using swirling flows is widely used in gas turbines and combustion. Although flashback is not generally a problem with natural gas combustion, there are some reports of flashback damage with existing gas turbines, whilst hydrogen enriched fuel blends cause concerns in this area. Thus, this paper describes a practical approach to study and avoid flashback in a pilot scale 100 kW tangential swirl burner. The flashback phenomenon is studied experimentally via the derivation of flashback limits for a variety of different geometrical conditions. A high speed camera is used to visualize the process and distinguish new patterns of avoidance. The use of a central fuel injector is shown to give substantial benefits in terms of flashback resistance. Conclusions are drawn as to mitigation technologies.

  13. Draft layout, containment and performance of the safety system of the European Supercritical Water-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schlagenhaufer, M.; Kohly, C.; Schulenberg, T. [Karlsruhe Inst. of Tech., Karlsruhe (Germany); Rothschmitt, S.; Bittermann, D. [AREVA NP GmbH, Erlangen (Germany)


    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. Close to the end of the project the technical results are translated into a draft layout of the HPLWR. The containment and safety system are being explained. Exemplarily, a depressurization event shows the capabilities of the safety system to sufficiently cool the reactor by means of a low pressure coolant injection system. (author)

  14. Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

    Energy Technology Data Exchange (ETDEWEB)

    Sponza, Delia Teresa, E-mail: [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)


    In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

  15. Firing in fluid beds and burners

    Energy Technology Data Exchange (ETDEWEB)

    Frandsen, F.; Lans, R. van der; Storm Pedersen, L.; Philbert Nielsen, H.; Aslaug Hansen, L.; Lin, W.; Johnsson, J.E.; Dam-Johansen, K.


    An investigation of the effect of co-firing straw and pulverized coal was performed. Based on experiments from pilot-scale and full-scale it was concluded that a higher fraction of straw in the fuel feedstock mixture results in lower NO and SO{sub 2} emissions. The lower NO emission was mainly due to the higher volatile content of the straw, which leads to lower stoichiometry in the gas phase and in subsequent suppression of NO{sub x} formation. This conclusion is consistent with experimental and modeling results for pure coal combustion. The effect of coal quality on NO emissions has been investigated with three coals of different characteristics in three furnaces: in the Funen power station, unit 7 (FVO7), the Midtkraft Studstrup power station, unit 4 (MKS4), and the Mitsui Babcock Energy Ltd (MBEL) test-rig. The MBEL test-rig was able to reproduce qualitatively the emissions from the MKS4 plant, and quantitatively the emissions from the FVO7 plant. The better agreement between the MBEL test-rig and FVO7 is presumed to be related to the existence of a large primary zone with a relatively low stoichiometry, diminishing the influence of near burner air and fuel mixing rate on the NO emissions. An engineering model has been developed for the prediction of NO emissions and burnout from pulverized fuel combustion in swirl burners. A simplified model for reduction of N{sub 2}O in CFBC has been developed, and simulation results are in good agreement with experimental data from a 12 MW{sub th} CFB-boiler. (EG) EFP-94. 108 refs.

  16. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)


    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  17. Self locking drive system for rotating plug of a nuclear reactor (United States)

    Brubaker, James E.


    This disclosure describes a self locking drive system for rotating the plugs on the head of a nuclear reactor which is able to restrain plug motion if a seismic event should occur during reactor refueling. A servomotor is engaged via a gear train and a bull gear to the plug. Connected to the gear train is a feedback control system which allows the motor to rotate the plug to predetermined locations for refueling of the reactor. The gear train contains a self locking double enveloping worm gear set. The worm gear set is utilized for its self locking nature to prevent unwanted rotation of the plugs as the result of an earthquake. The double enveloping type is used because its unique contour spreads the load across several teeth providing added strength and allowing the use of a conventional size worm.

  18. Combustion flame-plasma hybrid reactor systems, and chemical reactant sources (United States)

    Kong, Peter C


    Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

  19. Design of conduction cooling system for a high current HTS DC reactor (United States)

    Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun


    A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

  20. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Boustani Ehsan


    Full Text Available One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in many of abnormal situations. It is designed as much as practical based on neutron absorber solution injection into the existing core while the changes and interferences with the existing core structure are kept to a minimum. Core neutronic calculations were performed using MCNPX 2.6.0 and MTR_PC package for the current operational core equipped with the second shutdown system, and one experiment was conducted in the Tehran Research Reactor to test the neutronic calculations. A good agreement was seen between theoretical results and experimental ones. In addition, capability of the second shutdown system in the case of occurrence of design basis accident in the Tehran Research Reactor is demonstrated using PARET program.

  1. Investigation of the Combustion Stability of Methane-Air Mixture in Recuperative Burners of Different Geometries

    Directory of Open Access Journals (Sweden)

    Krainov Alexey


    Full Text Available The results of numerical investigations of 5.5% methane-air combustion stability in heat recuperative burners (counter flow burner, U-shape burner and Swiss-roll burner are presented in this paper. The investigation is carried out with the use of commercial CFD package Ansys-Fluent. The boundaries of combustion stability of 5.5% methane-air mixture depending on the gas flow rate at the inlet of the burners are determined.

  2. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)


    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  3. IR sensor for monitoring of burner flame; IR sensor foer oevervakning av braennarflamma

    Energy Technology Data Exchange (ETDEWEB)

    Svanberg, Marcus; Funkquist, Jonas; Clausen, Soennik; Wetterstroem, Jonas


    To obtain a smooth operation of the coal-fired power plants many power plant managers have installed online mass flow measurement of coal to all burners. This signal is used to monitor the coal mass flow to the individual burner and match it with appropriate amount of air and also to monitor the distribution of coal between the burners. The online mass flow measurement system is very expensive (approximately 150 kEUR for ten burners) and is not beneficial for smaller plants. The accuracy of the measurement and the sample frequency are also questionable. The idea in this project has been to evaluate a cheaper system that can present the same information and may also provide better accuracy and faster sample frequency. The infrared sensor is a cheap narrow banded light emission sensor that can be placed in a water cooed probe. The sensor was directed at the burner flame and the emitted light was monitored. Through calibration the mass flow of coal can be presented. Two measurement campaigns were performed. Both campaigns were carried out in Nordjyllandsverket in Denmark even though the second campaign was planned to be in Uppsala. Due to severe problems in the Uppsala plant the campaign was moved to Nordjyllandsverket. The pre-requisites for the test plant were that online measurement of coal flow was installed. In Nordjyllandsverket 4 out of 16 burners have the mass flow measurement installed. Risoe Laboratories has vast experiences in the IR technology and they provided the IR sensing equipment. One IR sensor was placed in the flame guard position just behind the flame directed towards the ignition zone. A second sensor was placed at the boiler wall directed towards the flame. The boiler wall position did not give any results and the location was not used during the second campaign. The flame-guard-positioned-sensor- signal was thoroughly evaluated and the results show that there is a clear correlation between the coal mass flow and the IR sensor signal. Tests were

  4. Nuclear reactor with makeup water assist from residual heat removal system (United States)

    Corletti, Michael M.; Schulz, Terry L.


    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  5. Oxygen transport membrane reactor based method and system for generating electric power (United States)

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan


    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  6. Municipal waste stabilization in a reactor with an integrated active and passive aeration system. (United States)

    Kasinski, Slawomir; Slota, Monika; Markowski, Michal; Kaminska, Anna


    To test whether an integrated passive and active aeration system could be an effective solution for aerobic decomposition of municipal waste in technical conditions, a full-scale composting reactor was designed. The waste was actively aerated for 5d, passively aerated for 35 d, and then actively aerated for 5d, and the entire composting process was monitored. During the 45-day observation period, changes in the fractional, morphological and physico-chemical characteristics of the waste at the top of the reactor differed from those in the center of the reactor. The fractional and morphological analysis made during the entire process of stabilization, showed the total reduction of organic matter measured of 82 wt% and 86 wt% at the respective depths. The reduction of organic matter calculated using the results of Lost of Ignition (LOI) and Total Organic Carbon (TOC) showed, respectively, 40.51-46.62% organic matter loss at the top and 45.33-53.39% in the center of the reactor. At the end of the process, moisture content, LOI and TOC at the top were 3.29%, 6.10% and 4.13% higher, respectively, than in the center. The results showed that application of passive aeration in larger scale simultaneously allows the thermophilic levels to be maintained during municipal solid waste composting process while not inhibiting microbial activity in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. Sewage treatment in integrated system of UASB reactor and duckweed pond and reuse for aquaculture. (United States)

    Mohapatra, D P; Ghangrekar, M M; Mitra, A; Brar, S K


    The performance of a laboratory-scale upflow anaerobic sludge blanket (UASB) reactor and a duckweed pond containing Lemna gibba was investigated for suitability for treating effluent for use in aquaculture. While treating low-strength sewage having a chemical oxygen demand (COD) of typically less than 200 mg/L, with an increase in hydraulic retention time (HRT) from 10.04 to 33.49 h, COD removal efficiency of the UASB reactor decreased owing to a decrease in organic loading rate (OLR) causing poor mixing in the reactor. However, even at the lower OLR (0.475 kg COD/(m3 x d)), the UASB reactor gave a removal efficiency of 68% for COD and 74% for biochemical oxygen demand (BOD). The maximum COD, BOD, ammonia-nitrogen and phosphate removal efficiencies of the duckweed pond were 40.77%, 38.01%, 61.87% and 88.57%, respectively. Decreasing the OLR by increasing the HRT resulted in an increase in efficiency of the duckweed pond for removal of ammonia-nitrogen and phosphate. The OLR of 0.005 kg COD/(m2 x d) and HRT of 108 h in the duckweed pond satisfied aquaculture quality requirements. A specific growth rate of 0.23% was observed for tilapia fish fed with duckweed harvested from the duckweed pond. The economic analysis proved that it was beneficial to use the integrated system of a UASB reactor and a duckweed pond for treatment of sewage.

  8. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun


    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  9. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun


    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  10. Radiotracer investigation on the measurement of residence time distribution in an ethyl acetate reactor system with a large recycle ratio. (United States)

    Datta, Arghya; Kumar Gupta, Raj; Goswami, Sunil; Kumar Sharma, Vijay; Bhunia, Haripada; Singh, Damandeep; Jagat Pant, Harish


    A radiotracer investigation was carried out on the measurement of residence time distribution (RTD) of process fluid in an industrial-scale ethyl acetate reactor system, which consists of two independent reactors with recirculation and connected in series with each other. Bromine-82 as ammonium bromide was used as the radiotracer for the RTD experiments at different operating conditions. The individual reactors and the overall reactor system were modelled using physically representative phenomenological models comprising of continuously stirred tank reactors (CSTRs). The results showed that the recirculation rate considerably affected the flow mixing behaviour and mean residence time of the process fluid in the reactor system. The results also showed that there was bypassing of the fluid in the first reactor that ranged from 12% to 22% and 40% dead volume at different operating conditions, whereas the second reactor behaved closely as an ideal CSTR. The results of the investigation can be used to optimise the process parameters and design new improved reactor systems for the production of ethyl acetate. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Study of DNA damage with a new system for irradiation of samples in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gual, Maritza R., E-mail: mrgual@instec.c [Instituto Superior de Tecnologias y Ciencias Aplicadas, InSTEC, Avenida Salvador Allende y Luaces, Quinta de Los Molinos, Plaza de la Revolucion, Havana, AP 6163 (Cuba); Milian, Felix M. [Universidade Estadual de Santa Cruz, UESC (Brazil); Deppman, Airton [Instituto de Fisica, Universidad de Sao Paulo, IF-USP, Rua do Matao, Travessa R, no. 187, Ciudade Universitaria, Butanta, CEP 05508-900, Sao Paulo (Brazil); Coelho, Paulo R.P. [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP (Brazil)


    In this paper, we report results of a quantitative analysis of the effects of neutrons on DNA, and, specifically, the production of simple and double breaks of plasmid DNA in aqueous solutions with different concentrations of free-radical scavengers. The radiation damage to DNA was evaluated by electrophoresis through agarose gels. The neutron and gamma doses were measured separately with thermoluminescent detectors. In this work, we have also demonstrated usefulness of a new system for positioning and removing samples in channel BH3 of the IEA-R1 reactor at the Instituto de Pesquisas Energeticas e Nucleares (Brazil) without necessity of interrupting the reactor operation.

  12. Design and development of a low-temperature reactor system for biorefining waste oil


    Pedersen, Hans Olav


    The background for this master’s thesis is the focus on bioenergy and biofuels at NMBU. This has, among others, resulted in a prototype of a small-scale biorefinery, which uses methanol and waste cooking oil to produce biodiesel. The purpose of this thesis is to develop a reactor system that serves as a platform for reactors to operate on and a technological alternative for a periodically on-site clean of catalysts. The purpose of the catalysts wash is to extend their lifetime, in order to ma...

  13. Automation of the radiation protection monitoring system in the RP-10 reactor; Automatizacion del sistema de monitoraje de radioproteccion en el reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Anaya G, Olgger; Castillo Y, Walter; Ovalle S, Edgar [Instituto Peruano de Energia Nuclear, Lima (Peru)


    During the reactor operation, it is necessary to carry out the radiological control in the different places of the reactor, in periodic form and to take a registration of these values. For it the radioprotection official, makes every certain periods, settled down in the procedures, to verify and to carry out the registration of those values in manual form of each one of the radiation monitors. For this reason it was carried out the design and implementation of an automatic monitoring system of radioprotection in the reactor. In the development it has been considered the installation of a acquisition data system for 27 radiation gamma monitors of the type Geiger Mueller, installed inside the different places of the reactor and in the laboratories where they are manipulated radioactive material, using as hardware the FieldPoint for the possessing and digitalization of the signs which are correspondents using the communication protocol RS-232 to a PC in which has settled a program in graphic environment that has been developed using the tools of the programming software LabWindows/CVI. Then, these same signs are sent 'on line' to another PC that is in the Emergency Center of Coordination to 500 m of the reactor, by means of a system of radiofrequency communication. (author)

  14. Overview of Progress on the EU DEMO Reactor Magnet System Design

    NARCIS (Netherlands)

    Zani, L.; Bayer, C.; biancolini, M.E.; Bonifetto, R.; Nijhuis, Arend; Yagotintsev, K.


    The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In

  15. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.


    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  16. Evaluation of the dual digestion system 2: operation and performance of the pure oxygen aerobic reactor

    CSIR Research Space (South Africa)

    Messenger, JR


    Full Text Available In a comprehensive study of the performance of a full-scale (45 m3) pure oxygen autothermal thermophilic aerobic reactor of a sewage sludge dual digestion system, it was found that: Biological heat generation rate was directly proportional...

  17. Reactor-Capaсitor Device for Flexible Link Between Non-Synchronous Power Systems

    Directory of Open Access Journals (Sweden)

    Bosneaga V.


    Full Text Available In present flexible interconnections for transmission of required active power between different power systems is used, as a rule, so-called DC back-to-back link. The aim of this work is the investigation of proposed reactor-capacitor device for flexible connection of asynchronously alternating current power systems with the same nominal values of frequencies for parallel operation. The reactor-capacitor device was elaborated. The installation develops the idea of controlled reactor alternating current link, and provides reactive power balance in the unit and needed value of the output voltage module. The basic characteristics of reactor-capacitor device for controlled power transmission were investigated. Analytical expressions for device elements parameters were derived. These ensure necessary ratio of voltages modules of linked power systems and reactive power balance of the device at circular output voltage vector rotation for a given load admittance. Obtained parameters ensure constant active power flow between linked asynchronously power systems and device reactive power internal balance.

  18. New approach to control the methanogenic reactor of a two-phase anaerobic digestion system

    Energy Technology Data Exchange (ETDEWEB)

    Sachs, J. von; Meyer, U.; Rys, P.; Feitkenhauer, H. [ETH Zurich (Switzerland). Laboratorium fuer Technische Chemie


    A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor. (author)

  19. Innovative inspection system for reactor pressure vessels; Innovative Pruefsysteme fuer Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, K.; Trautmann, H.


    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [Deutsch] Die vorgestellten kleinen, flexiblen und modernen Schwimmsysteme (DELPHIN-Manipulatoren und MIDAS-U-Boote) sind innovative Systeme fuer die Reduzierung der Aufwaende und Zeit zur Pruefung des Reaktordruckbehaelters und damit zur Reduktion der Revisionszeiten der Reaktoranlagen. (orig.)

  20. Ion transport membrane reactor systems and methods for producing synthesis gas (United States)

    Repasky, John Michael


    Embodiments of the present invention provide cost-effective systems and methods for producing a synthesis gas product using a steam reformer system and an ion transport membrane (ITM) reactor having multiple stages, without requiring inter-stage reactant injections. Embodiments of the present invention also provide techniques for compensating for membrane performance degradation and other changes in system operating conditions that negatively affect synthesis gas production.

  1. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR). (United States)

    Tawfik, A; El-Kamah, H


    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

  2. Lasers and power systems for inertial confinement fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stark, E.E. Jr.


    After discussing the role of lasers in ICF and the candidate lasers, several important areas of technology requirements are discussed. These include the beam transport system, the pulsed power system and the gas flow system. The system requirements, state of the art, as well as needs and prospects for new technology developments are given. Other technology issues and promising developments are described briefly.

  3. Simulation tools for the design of natural gas domestic burners

    Energy Technology Data Exchange (ETDEWEB)

    Hilka, M. [DEG Gaz de France, Saint Denise la Plaine (France). Direction de la Recherche; Quilichini, V.; Gicquel, O.; Darabiha, N. [Laboratoire E.M2.C., Ecole Centrale Paris, CNRS, Chatenay-Malabry (France)


    The design of domestic burners crucially depends on the availability of tools taking into account complex interactions between flame chemistry, heat transfer and fluid flow. A very promising approach is therefore the development of modern simulation tools incorporating appropriate physical models that enable the predicition of flame stability and pollutant formation in practical devices. Given the complex, 3D geometry of practical burners, we decided to adapt the commercially available, general purpose CFD-code ESTET to the simulation of combustion in domestic burners. This has been achieved through the implementation of a complex chemical kinetics library (BISCUIT) within the CFD code and an adaptation of the graphical user interface. The resulting tool is capable to predict partially premixed flames that characterize domestic burners, as well as the formation of pollutants such as NO{sub x} and has been carefully validated against experimental data obtained for a model burner. Computational ressources required for multi-dimensional burner configurations are standard UNIX workstations. Computing time typically varies from 3 h to 150 h, depending on the physical models used. (orig.)

  4. Design process and instrumentation of a low NOx wire-mesh duct burner for micro-cogeneration unit

    Energy Technology Data Exchange (ETDEWEB)

    Ramadan, O.B.; Gauthier, J.E.D. [Carleton Univ., Ottawa, ON (Canada). Dept. of Mechanical and Aerospace Engineering; Hughes, P.M.; Brandon, R. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Energy Technology Centre


    Air pollution and global climate change have become a serious environmental problem leading to increasingly stringent government regulations worldwide. New designs and methods for improving combustion systems to minimize the production of toxic emissions, like nitrogen oxides (NOx) are therefore needed. In order to control smog, acid rain, ozone depletion, and greenhouse-effect warming, a reduction of nitrogen oxide is necessary. One alternative for combined electrical power and heat generation (CHP) are micro-cogeneration units which use a micro-turbine as a prime mover. However, to increase the efficiencies of these units, micro-cogeneration technology still needs to be developed further. This paper described the design process, building, and testing of a new low NOx wire-mesh duct burner (WMDB) for the development of a more efficient micro-cogeneration unit. The primary goal of the study was to develop a practical and simple WMDB, which produces low emissions by using lean-premixed surface combustion concept and its objectives were separated into four phases which were described in this paper. Phase I involved the design and construction of the burner. Phase II involved a qualitative flow visualization study for the duct burner premixer to assist the new design of the burner by introducing an efficient premixer that could be used in this new application. Phase III of this research program involved non-reacting flow modeling on the burner premixer flow field using a commercial computational fluid dynamic model. In phase IV, the reacting flow experimental investigation was performed. It was concluded that the burner successfully increased the quantity and the quality of the heat released from the micro-CHP unit and carbon monoxide emissions of less than 9 ppm were reached. 3 refs., 3 figs.

  5. Design requirements of instrumentation and control systems for next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.

  6. HEMERA: a 3D coupled core-plant system for accidental reactor transient simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, G.B.; Fouquet, F.; Dubois, F. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France); Le Pallec, J.C.; Richebois, E.; Hourcade, E.; Poinot-Salanon, C.; Royer, E. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S), 91 - Gif sur Yvette (France)


    In the framework of their collaboration to develop a system to study reactor transients in safety-representative conditions, IRSN (Radioprotection and Nuclear Safety Institute) and Cea have launched the development of a fully coupled 3-dimensional computational chain, called HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed of APOLLO-2, CRONOS-2 and FLICA-4 codes, and the system code CATHARE. It includes cross sections generation, steady-state, depletion and transient computation capabilities in a consistent approach. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermohydraulics. Currently Control Rod Ejection (RIA) and Main Steam Line Break (MSLB) accidents are investigated. The HEMERA system is presently applied to French PWR. The present paper outlines the main physical phenomena to be accounted for in such a coupled computational chain with significant time and space effects.

  7. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail:, E-mail:, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)


    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  8. Study of reactor Brayton power systems for nuclear electric spacecraft (United States)


    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  9. Technical report on implementation of reactor internal 3D modeling and visual database system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

  10. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)


    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  11. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100


    Zhao, Pengcheng; Shi, Kangli; Li, Shuzhou; Feng, Jingchao; Chen, Hongli


    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  12. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)


    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  13. The under-critical reactors physics for the hybrid systems; La physique des reacteurs sous-critiques des systemes hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Schapira, J.P. [Institut de Physique Nucleaire, IN2P3/CNRS 91 - Orsay (France); Vergnes, J. [Electricite de France, EDF, Direction des Etudes et Recherches, 75 - Paris (France); Zaetta, A. [CEA/Saclay, Direction des Reacteurs Nucleaires, DRN, 91 - Gif-sur-Yvette (France)] [and others


    This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

  14. Scaleable, High Efficiency Microchannel Sabatier Reactor Project (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  15. The Dynamic Anaerobic Reactor & Integrated Energy System (DARIES) model: model development, validation, and sensitivity analysis. (United States)

    Brouwer, A F; Grimberg, S J; Powers, S E


    The Dynamic Anaerobic Reactor & Integrated Energy System (DARIES) model has been developed as a biogas and electricity production model of a dairy farm anaerobic digester system. DARIES, which incorporates the Anaerobic Digester Model No. 1 (ADM1) and simulations of both combined heat and power (CHP) and digester heating systems, may be run in either completely mixed or plug flow reactor configurations. DARIES biogas predictions were shown to be statistically coincident with measured data from eighteen full-scale dairy operations in the northeastern United States. DARIES biogas predictions were more accurate than predictions made by the U.S. AgSTAR model FarmWare 3.4. DARIES electricity production predictions were verified against data collected by the NYSERDA DG/CHP Integrated Data System. Preliminary sensitivity analysis demonstrated that DARIES output was most sensitive to influent flow rate, chemical oxygen demand (COD), and biodegradability, and somewhat sensitive to hydraulic retention time and digester temperature.

  16. Hybrid model predictive control applied to switching control of burner load for a compact marine boiler design

    DEFF Research Database (Denmark)

    Solberg, Brian; Andersen, Palle; Maciejowski, Jan


    This paper discusses the application of hybrid model predictive control to control switching between different burner modes in a novel compact marine boiler design. A further purpose of the present work is to point out problems with finite horizon model predictive control applied to systems for w...

  17. The Pressure Relief System Design for Industrial Reactors

    Directory of Open Access Journals (Sweden)

    Iztok Hace


    Full Text Available A quick and simple approach for reactor—emergency relief system design—for runaway chemical reactions is presented. A cookbook for system sizing with all main characteristic dimensions and parameters is shown on one realistic example from process industry. System design was done based on existing theories, standards, and correlations obtained from the literature, which were implemented for presented case. A simple and effective method for emergency relief system is shown, which may serve as an example for similar systems design. Obtained results may contribute to better understanding of blow down system frequently used in industrial plants, for increasing safety, decreasing explosion damage, and alleviating the ecological problems together with environmental pollution in case of industrial accidents.

  18. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    OHara J. M.; Higgins, J.; DAgostino, A.


    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  19. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W


    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  20. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)


    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  1. Flashback Analysis in Tangential Swirl Burners

    Directory of Open Access Journals (Sweden)

    Valera-Medina A.


    Full Text Available Premixed lean combustion is widely used in Combustion Processes due to the benefits of good flame stability and blowoff limits coupled with low NOx emissions. However, the use of novel fuels and complex flows have increased the concern about flashback, especially for the use of syngas and highly hydrogen enriched blends. Thus, this paper describes a combined practical and numerical approach to study the phenomenon in order to reduce the effect of flashback in a pilot scale 100 kW tangential swirl burner. Natural gas is used to establish the baseline results and effects of different parameters changes. The flashback phenomenon is studied with the use of high speed photography. The use of a central fuel injector demonstrates substantial benefits in terms of flashback resistance, eliminating coherent structures that may appear in the flow channels. The critical boundary velocity gradient is used for characterization, both via the original Lewis and von Elbe formula and via analysis using CFD and investigation of boundary layer conditions in the flame front.

  2. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems (United States)

    Hançerliogulları, Aybaba; Cini, Mesut


    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  3. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail:


    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  4. Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation (United States)

    Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François


    A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

  5. Application of the BEACON-TSM system to the operation of PWR reactors; Aplicacion del sistema Beacon TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.


    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  6. Influence of burner form and pellet type on domestic pellet boiler performance (United States)

    Rastvorov, D. V.; Osintsev, K. V.; Toropov, E. V.


    The study presents combustion and emission results obtained using two serial pellet boilers of the same heating capacity 40 kW. These boilers have been designed by producers for domestic conditions of exploitation. The principal difference between boilers was the type of the burner. The study concerns the efficiency and ecological performance difference between burners of circular and rectangular forms. The features of the combustion process in both types of burners were studied when boiler operated with different sorts of pellets. The results suggest that the burner of circular form excels the rectangular form burner. However, there is some difference of NOx emission between circular and rectangular burners.

  7. Coal Particle Flow Patterns for O2 Enriched, Low NOx Burners

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Sinclair Curtis


    This project involved a systematic investigation examining the effect of near-flame burner aerodynamics on standoff distance and stability of turbulent diffusion flames and the resultant NO{sub x} emissions from actual pulverized coal diffusion flames. Specifically, the scope of the project was to understand how changes in near-flame aerodynamics and transport air oxygen partial pressure can influence flame attachment and coal ignition, two properties essential to proper operation of low NO{sub x} burners. Results from this investigation utilized a new 2M tall, 0.5m in diameter combustor designed to evaluate near-flame combustion aerodynamics in terms of transport air oxygen partial pressure (Po{sub 2}), coal fines content, primary fuel and secondary air velocities, and furnace wall temperature furnish insight into fundamental processes that occur during combustion of pulverized coal in practical systems. Complementary cold flow studies were conducted in a geometrically similar chamber to analyze the detailed motion of the gas and particles using laser Doppler velocimetry. This final technical report summarizes the key findings from our investigation into coal particle flow patterns in burners. Specifically, we focused on the effects of oxygen enrichment, the effect of fines, and the effect of the nozzle velocity ratio on the resulting flow patterns. In the cold flow studies, detailed measurements using laser Doppler velocimetry (LDV) were made to determine the details of the flow. In the hot flow studies, observations of flame stability and measurements of NO{sub x} were made to determine the effects of the flow patterns on burner operation.

  8. Robotic dismantlement systems at the CP-5 reactor D&D project.

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, L. S.


    The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D&D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D&D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building

  9. Development of a non-premixed radiant burner. Experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Andersen, P.; Myken, A.N.; Rasmussen, N.B.


    The objective was to develop and test a non-premixed radiant burner. A burner concept, pro-type 1, was developed with special attention to minimise emissions and foam peak-temperatures. During the test of the prototype it became evident that the construction was not feasible, because the ceramic gas distribution tubes eventually were blocked by soot. Experiments to investigate if the gas would crack conducted prior to the construction of the burner did not indicate that any problems would occur. It can therefore be concluded that the experiments did not simulate burner conditions adequately. Alternative prototypes in which the gas is not heated prior to injection into the combustion chamber have been established. The concept operated satisfactorily without preheating of the combustion air, although NO{sub x}-emissions were high. The measured process efficiencies were superior to previous results for different kinds of surface burners. When the combustion air was preheated to 400 deg. C, the foam sections broke down. The experimental results can be summarised in the following conclusions: The developed prototypes can not be operated with combustion air preheated to 400 deg. C or higher; A relative improvement of the process efficiency by 22% has been observed when the combustion air is preheated to 400 deg. C; The NO{sub x}-emissions increase significantly and much more than the process efficiency when the combustion air is preheated; The process efficiency obtained with prototype 5 is better than previously investigated surface burners, especially at high loads. Possible means to improve durability, efficiency and emission level for both burner concepts are suggested. These include cooling of the gas in prototype 1, coating of the downstream side of the foam section to improve the radiant efficiency and multistep combustion. (EHS)

  10. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices (United States)

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R


    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  11. Application of CALPUFF to PM10 emissions from beehive burners in British Columbia

    Energy Technology Data Exchange (ETDEWEB)

    Ciccone, A.D. [Jacques Whitford and Associates Limited, Vancouver, BC (Canada); Waddell, G. [Canadian Forest Products Ltd., Prince George, BC (Canada)


    The complex local topography of the Bulkley Valley in the British Columbia interior greatly influences the local meteorology and climatology. The communities of Smithers and Houston which are located in the valley are hosts to three mills which operate conical burners for waste disposal and which define the extent of airshed. The CALMET/CALPUFF modelling system was chosen as a means to evaluate the contribution of the burners to the local airshed. CALPUFF was chosen because of the combined conditions of complex terrain and low wind speed in the region. Since MM5 gridded meteorological data was available from the BC Ministry of Environment to initialize the wind fields for CALMET in 1995, modelling was conducted in that year. CALPUFF provided 24-hour PM10 ground level concentrations over a 54 km by 72 km range. This included monitoring stations in the airshed. The impact from the conical burners was found to be low compared to the monitoring data which was collected. However, it was determined that the model was able to describe hourly changes in ambient PM10 levels, which reflected the hourly monitoring station data. The region is now equipped with a modelling platform that can be used to help in air pollution source appointment as well as for the management general air quality.

  12. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo


    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  13. Development of fiber-delivered laser peening system to prevent stress corrosion cracking of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Sano, Y.; Kimura, M.; Yoda, M.; Mukai, N.; Sato, K.; Uehara, T.; Ito, T.; Shimamura, M.; Sudo, A.; Suezono, N. [Toshiba Corp., Yokohama (Japan)


    The authors have developed a system to deliver water-penetrable intense laser pulses of frequency-doubled Nd:YAG laser through optical fiber. The system is capable of improving a residual stress on water immersed metal material remotely, which is effective to prevent the initiation of stress corrosion cracking (SCC) of reactor components. Experimental results showed that a compressive residual stress with enough amplitude and depth was built in the surface layer of type 304 stainless steel (SUS304) by irradiating laser pulses through optical fiber with diameter of 1 mm. A prototype peening head with miniaturized dimensions of 88 mm x 46 mm x 25 mm was assembled to con-firm the accessibility to the heat affected zone (HAZ) along weld lines of a reactor core shroud. The accessibility was significantly improved owing to the flexible optical fiber and the miniaturized peening head. The fiber delivered system opens up the possibility of new applications of laser peening. (author)

  14. Application of fault detection and identification (FDI) techniques in power regulating systems of nuclear reactors (United States)

    Roy, K.; Banavar, R. N.; Thangasamy, S.


    Application of failure detection and identification (FDI) algorithms have essentially been limited to identification of a global fault in the system, and no further attempts have been made to locate subcomponent faults for root cause analysis. This paper presents Kalman filter-based methods for FDI in power regulating systems of nuclear reactors. The attempt here is to explain how the behavior of the states, residues, and covariances can be interpreted to identify subcomponent faults. An alternative to the Kalman filter-the risk-sensitive filter-is also introduced. Comparison of its performance with the Kalman filter-based FDI algorithms is studied. All simulation studies have been carried out on postulated faults in the power regulating system of heavy water moderated, low pressure vertical tank-type research reactors.

  15. A fancy eco-compatible wastewater treatment system: Green Bio-sorption Reactor. (United States)

    Zhao, Yaqian; Liu, Ranbin; Zhao, Jinhui; Xu, Lei; Sibille, Caroline


    A novel concept was proposed and preliminarily investigated by embedding alum sludge-based constructed wetland into conventional activated sludge system in terms of Green Bio-sorption Reactor (GBR). This novel GBR inherited the aesthetic value of constructed wetland and owned the robust phosphorus (P) adsorption along with the benefit of carriers' addition (dewatered alum sludge). The preliminary demonstration was conducted in a lab-scale sequencing batch reactor (SBR) system without biological phosphorus removal process. The novel process achieved averagely 96%, 99% and 90% for BOD, TP and TN removal with piggery wastewater as influent, demonstrating for the first time of its promising performance. Moreover, the coexistence of biofilm and suspended sludge also achieved 55-88% simultaneous nitrification and denitrification efficiency, higher than biofilm only. Overall, alum sludge-based GBR could achieve reliable pollutants removal and provides a novel and sustainable pathway to upgrade conventional activated sludge system. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Startup thaw concept for the SP-100 space reactor power system (United States)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.


    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.


    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory


    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  18. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing (United States)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.


    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  19. Model Reduction Using Proper Orthogonal Decomposition and Predictive Control of Distributed Reactor System

    Directory of Open Access Journals (Sweden)

    Alejandro Marquez


    Full Text Available This paper studies the application of proper orthogonal decomposition (POD to reduce the order of distributed reactor models with axial and radial diffusion and the implementation of model predictive control (MPC based on discrete-time linear time invariant (LTI reduced-order models. In this paper, the control objective is to keep the operation of the reactor at a desired operating condition in spite of the disturbances in the feed flow. This operating condition is determined by means of an optimization algorithm that provides the optimal temperature and concentration profiles for the system. Around these optimal profiles, the nonlinear partial differential equations (PDEs, that model the reactor are linearized, and afterwards the linear PDEs are discretized in space giving as a result a high-order linear model. POD and Galerkin projection are used to derive the low-order linear model that captures the dominant dynamics of the PDEs, which are subsequently used for controller design. An MPC formulation is constructed on the basis of the low-order linear model. The proposed approach is tested through simulation, and it is shown that the results are good with regard to keep the operation of the reactor.

  20. Advances in high rate anaerobic treatment: staging of reactor systems.

    NARCIS (Netherlands)

    Lier, van J.B.; Zee, van der F.P.; Tan, N.C.G.; Rebac, S.; Kleerebezem, R.


    Anaerobic wastewater treatment (AnWT) is considered as the most cost-effective solution for organically polluted industrial waste streams. Particularly the development of high-rate systems, in which hydraulic retention times are uncoupled from solids retention times, has led to a world-wide

  1. Reactivity Monitoring of Accelerator-Driven Nuclear Reactor Systems

    NARCIS (Netherlands)

    Uyttenhove, W.


    This thesis provides a methodology and set-up of a reactivity monitoring tool for Accelerator-Driven Systems (ADS). The reactivity monitoring tool should guarantee the operation of an ADS at a safe margin from criticality. Robustness is assured in different aspects of the monitoring tool: the choice

  2. Development of Environmentally-Assisted Fatigue Monitoring System for Advanced Power Reactors (APR1400)

    Energy Technology Data Exchange (ETDEWEB)

    Park, June Soo; Kim, Yeon Jeong; Kang, Sun Yeh; Yoon, Ki Seok; Choi, Taek Sang [KEPCO-E and C, Daejeon (Korea, Republic of)


    This paper introduces an EAF monitoring system developed for Shin-Kori Nuclear Power Plant (NPP), Units 3 and 4 which are the first two reactors of the APR1400 model. The EAF monitoring system has been developed for Shin-Kori NPP, Units 3 and 4, and is ready for an application for the plant lifetime. It is expected that the plant fatigue management can be effectively fulfilled, and the structural integrity of the critical components assured by an implementation of the fatigue monitoring system from the beginning of the lifetime. When fatigue analyses including the effects of the Light-Water Reactor (LWR) environment are applicable, plant designers address the environmentally-assisted fatigue (EAF) for Class 1 reactor pressure boundary components. The environment factor (F{sub en}) method has been endorsed by the U. S. Nuclear Regulatory Commission for evaluating fatigue analyses to address the environmental effects, and this method considers four major variables in addition to the traditional air-fatigue analyses: Material temperature, dissolved oxygen content of coolant, sulfur (S) content of material, and strain rate at the material points of interest. APR1400 nuclear power plants are designed to the requirements of the enhanced plant safety, availability and performance criteria for a 60 year design life. To better manage the material degradation and structural integrity of the pressure boundary components, a fatigue monitoring system has been developed for APR1400 NPPs, which is capable to monitor the EAF damage during the plant lifetime.

  3. RSYST: an integrated modular system with a data basis, for automated calculation of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ruehle, Roland [Univ. of Stuttgart (Germany)


    The integrated modular system RSYST was developed to offer the engineer and physicist a simpler method for carrying out the layout calculations for nuclear reactors. The system consists of a data basis, a control section, a data basis monitoring system, as well as an unlimited number of modules. The data transfer between individual modules is done through the data basis monitoring program by way of the central data basis. Control words of the input permit the user to control the process of any desired modules. Each module can have flexible data input to it from the data basis. By use of special modules, logical branches and loops can be carried out. The system was implemented on a CDC 6600 and partly on an IBM 360/75. At this moment, it includes 45,000 FORTRAN statements and 120 control words. Project calculations have been successfully carried out with the aid of RSYST for over three years. At this time, in addition to the general modules there exists primarily modules for reactor statistics calculations, burn-up calculations, and shielding calculations, and for the production of group constants. A start has been made to include problems of heat conduction, thermal hydraulics, reactor safety, control technology, and loop dynamics. (auth)

  4. Pressure suppression containment system for boiling water reactor (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.


    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  5. An advanced extruder-feeder biomass liquefaction reactor system (United States)

    White, Don H.; Wolf, D.; Davenport, G.; Mathews, S.; Porter, M.; Zhao, Y.


    A unique method of pumping concentrated, viscous biomass slurries that are characteristic of biomass direct liquefaction systems was developed. A modified single-screw extruder was shown to be capable of pumping solid slurries as high as 60 weight percent wood flour in wood oil derived vacuum bottoms, as compared to only 10 to 20 weight percent wood flour in wood oil in conventional systems. During the period August, 1985 to April, 1987, a total of 18 experimental continuous biomass liquefaction runs were made using white birch feedstock. Good operability with feed rates up to 30 lb/hr covering a range of carbon monoxide, sodium carbonate catalyst, pressures from 800 to 3000 psi and temperatures from 350 C to 430 C was achieved. Crude wood oils containing 6 to 10 weight percent residual oxygen were obtained. Other wood oil characteristics are reported.

  6. Progress in hardware development for the SAFE heatpipe reactor system (United States)

    Ring, P. J.; Sayre, E. D.; van Dyke, Melissa; Houts, Mike


    Advanced Methods & Materials Company (AMM) previously fabricated the stainless steel modules for the SAFE 30 system. These earlier modules consisting of five fuel pins surrounding a heat pipe, were brazed together using a tricusp insert in the gaps between tubes to ensure maximum braze coverage. It was decided that if possible the next generations of modules, both stainless steel and refractory alloy, would be diffusion bonded together using a Hot Issostatic Pressing (HIP) process. This process was very successfully used in producing the bonded rhenium Nb-lZr fuel cladding and the heat exchanger for the SP-100 Nuclear Space System Ref. 1 & 2. In addition AMM have since refined the technology enabling them to produce very high temperature rocket thrust chambers. Despite this background the complex geometry required for the SAFE module was quite challenging. It was necessary to develop a method which could be applied for both stainless steel and refractory alloy systems. In addition the interstices between tubes had to be completely filled with the tricusp insert to avoid causing distortion of the tube shape during HIPing and provide thermal conductivity from the fuel tubes to the heat pipes. Nevertheless it was considered worth the effort since Hot Isostatic Pressing, if successful, will produce an assembly with the heat pipe completely embedded within the module such that the diffusion bonded assembly has the thermal conduction and strength equivalent to a solid structure. .

  7. System for Coupling an IEC Reactor to Ion Thrusters (United States)

    Webber, Jason; Burton, Rodney; Momoto, Hiromu; Miley, George; Richardson, Nathan


    A conceptual design for an electric-thruster-driven space ship using a D-He3 fueled Inertial Electrostatic Confinement (IEC) fusion power unit was recently developed [1]. This propulsion system uses a bank of modified NSTAR-type krypton ion thrusters (specific impulse of 16,000 sec.) giving a total thrust of 1020 N. The thrust time for a typical outer planet mission ( e.g. Jupiter) with a delta-V of 50,000 m/s is then 200 days. A key component of this concept is a traveling wave direct energy converter that converts the kinetic energy of 14-MeV fusion reaction product protons to high voltage (about 1 MV) DC electrical output. A unique step-down transformer and rectifier system condition this output for use in the ion thrusters. Details of these components, the NSTAR-thruster modifications plus a magnetic hexa-pole collimator designed to guide the emitted protons into the traveling wave converter will be described. This advanced electric thruster design offers a very high power-to-weight ratio system that is crucial for deep space propulsion. [1] George H. Miley, Hiromu Momota, R. Burton, N.Richardson, M. Coventry, and Y. Shaban, IEC Based D-He3 Fusion for Space Propulsion, Trans Am. Nuclear Society, Annual Meeting, Hollywood, FL, June 2002.

  8. Reviewing real-time performance of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Preckshot, G.G. [Lawrence Livermore National Lab., CA (United States)


    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  9. Applying and adapting the Swedish regulatory system for decommissioning to nuclear power reactors - The regulator's perspective. (United States)

    Amft, Martin; Leisvik, Mathias; Carroll, Simon


    Half of the original 13 Swedish nuclear power reactors will be shut down by 2020. The decommissioning of these reactors is a challenge for all parties involved, including the licensees, the waste management system, the financing system, and the Swedish Radiation Safety Authority (SSM). This paper presents an overview of the Swedish regulations for decommissioning of nuclear facilities. It describes some of the experiences that SSM has gained from the application of these regulations. The focus of the present paper is on administrative aspects of decommissioning, such as SSM's guidelines, the definition of fundamental concepts in the regulatory framework, and a proposed revision of the licensing process according to the Environmental Act. These improvements will help to streamline the administration of the commercial nuclear power plant decommissioning projects that are anticipated to commence in Sweden in the near future. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri


    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  11. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)



    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  12. Priorities in the development of nuclear constants support system for reactor and shielding calculations

    Directory of Open Access Journals (Sweden)

    G.N. Manturov


    Development of the integral unified nuclear data support system and its implementation in the calculation codes will ensure not only the unification of the procedure for nuclear data preparation, which will allow enhancing reliability of their verification, but, as well, will enhance accuracy and reliability of calculation prediction of all the most important characteristics of the reactors under design, will ensure their licensing compliance, competitiveness and independence from foreign products.

  13. On the implementation of new technology modules for fusion reactor systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Franza, F., E-mail: [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Boccaccini, L.V.; Fisher, U. [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Gade, P.V.; Heller, R. [Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany)


    Highlights: • At KIT a new technology modules for systems code are under development. • A new algorithm for the definition of the main reactor's components is defined. • A new blanket model based on 1D neutronics analysis is described. • A new TF coil stress model based on 3D electromagnetic analysis is described. • The models were successfully benchmarked against more detailed models. - Abstract: In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished. In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.

  14. Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard Thomas [ORNL


    In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system. Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established infrastructures

  15. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail:, E-mail:, E-mail:, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear


    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  16. Convective wave front locking for a reaction-diffusion system in a conical flow reactor

    DEFF Research Database (Denmark)

    Kuptsov, P.V.; Kuznetsov, S.P.; Knudsen, Carsten


    We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute at the downstr......We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute...

  17. Modeling of adsorber/desorber/catalytic reactor system for ethylene oxide removal

    Directory of Open Access Journals (Sweden)



    Full Text Available The removal of ethylene oxide (EtO in a combined system adsorber/desorber/catalytic reactor has been investigated. The combined system was a modified draft tube spouted bed reactor loaded with Pt/Al2O3 catalyst. The annular region was divided into two sectons, the “hot” section contained about 7 % of catalyst and it behaved as a desorber and catalytic incinerator, while the “cold” section, with the rest of the catalyst, behaved as a sorber. The catalyst particles were circulated between the two sections by use of a draft tube riser. The Computational Fluid Dynamics (CFD program package FLUENT was used for simulations of the operation of the combined system. In addition, a one-dimensional numerical model for the operation of the packed bed reactor was compared with the corresponding FLUENT calculations. The results of the FLUENT simulations are in very good agreement with the experimental observations, as well as with the results of the one-dimensional numerical simulations.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  19. A survey of commercially available manipulators, end-effectors, and delivery systems for reactor decommissioning activities

    Energy Technology Data Exchange (ETDEWEB)

    Henley, D.R. [Argonne National Lab., IL (United States); Litka, T.J. [Advanced Consulting Group, Chicago, IL (United States)


    Numerous nuclear facilities owned by the U.S. Department of Energy (DOE) are under consideration for decommissioning. Currently, there are no standardized, automated, remote systems designed to dismantle and thereby reduce the size of activated reactor components and vessels so that they can be packaged and shipped to disposal sites. Existing dismantling systems usually consist of customized, facility-specific tooling that has been developed to dismantle a specific reactor system. Such systems have a number of drawbacks. Generally, current systems cannot be disassembled, moved, and reused. Developing and deploying the tooling for current systems is expensive and time-consuming. In addition, the amount of manual work is significant because long-handled tools must be used; as a result, personnel are exposed to excessive radiation. A standardized, automated, remote system is therefore needed to deliver the tooling necessary to dismantle nuclear facilities at different locations. Because this system would be reusable, it would produce less waste. The system would also save money because of its universal design, and it would be more reliable than current systems.

  20. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail:, E-mail:, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)


    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  1. Compact Green's Function for a Generic Rijke Burner

    Directory of Open Access Journals (Sweden)

    P. R. Murray


    Full Text Available A theoretical examination is made of the thermo-acoustic properties of a Rijke burner of large aspect ratio rectangular cross-section. Such a generic device has been proposed by Kok et al. (2009 paper presented at the 16th International Congress on Sound & Vibration to make canonical studies of combustion instabilities. An aeroacoustic Green's function is derived which permits the sound pressure produced by arbitrary thermal and vortex sources within the burner to be calculated by convolution. The Green's function corresponds to the potential flow sound field produced by an impulsive point source; its calculation taking account of flame-holder geometry is facilitated by use of the Schwarz-Christoffel transformation. The transformation is performed numerically to accommodate complex burner geometry and validated by comparison with an alternative procedure involving the direct numerical integration of Laplace's equation.

  2. Advanced Fusion Reactors for Space Propulsion and Power Systems (United States)

    Chapman, John J.


    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  3. Assessment of the reliability of neutronic parameters of Ghana Research Reactor-1 control systems

    Energy Technology Data Exchange (ETDEWEB)

    Amponsah-Abu, E.O., E-mail: [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Gbadago, J.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Akaho, E.H.K.; Akoto-Bamford, S. [School of Nuclear and Allied Sciences, University of Ghana (Ghana); Gyamfi, K.; Asamoah, M.; Baidoo, I.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana)


    Highlights: • The reliability of neutronics parameters of GHARR-I was assessed. • The reactor was operated at different power levels of 5–30 kW. • The pre-set flux was compared with the flux in the inner irradiation site. • Decrease in the core reactivity caused difference in flux on the meters and site. • Neutronic parameters become reliable when operation is done at reactivity of 4 mk. - Abstract: The Ghana Research Reactor-1 (GHARR-1) has been in operation for the past 19 years using a Micro-Computer Closed Loop System (MCCLS) and Control Console (CC) as the control systems. The two control systems were each coupled separately with a micro-fission chamber to measure the current pulses of the neutron fluxes in the core at excess reactivity of 4 mk. The MCCLS and CC meter readings at a pre-set flux of 5.0 × 10{sup 11} n/cm{sup 2} s were 6.42 × 10{sup 11} n/cm{sup 2} s and 5.0 × 10{sup 11} n/cm{sup 2} s respectively. Due to ageing and obsolescence, the MCCLS and some components that control the sensitivity and the reading mechanism of the meters were replaced. One of the fission chambers was also removed and the two control systems were coupled to one fission chamber. The reliability of the neutronic parameters of the control systems was assessed after the replacement. The results showed that when the reactor is operated at different power levels of 5–30 kW using one micro-fission chamber, the pre-set neutron fluxes at the control systems is 1.6 times the neutron fluxes obtained using a flux monitor at the inner irradiation site two of the reactor. The average percentage deviations of the obtained fluxes from the pre-set values of 1.67 × 10{sup 11}–1.0 × 10{sup 12} n/cm{sup 2} s were 36.5%. This compares very well with the decrease in core excess reactivity of 36.3% of the nominal value of 4 mk, after operating the reactor at critical neutron flux of 1.0 × 10{sup 9} n/cm{sup 2} s.

  4. Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O.


    Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type ({nu} = 10{sup -8}/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the {gamma}-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A{sup 41} in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ed Gorski; Dennis Harrell; Finis Southworth


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

  6. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.


    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  7. Opposed Jet Burner Approach for Characterizing Flameholding Potentials of Hydrocarbon Scramjet Fuels (United States)

    Pellett, Gerald L.; Convery, Janet L.; Wilson, Lloyd G.


    Opposed Jet Burner (OJB) tools have been used extensively by the authors to measure Flame Strength (FS) extinction limits of laminar H2/N2 air and (recently) hydrocarbon (HC) air Counterflow Diffusion Flames (CFDFs) at one atm. This paper details normalization of FSs of N2- diluted H2 and HC systems to account for effects of fuel composition, temperature, pressure, jet diameter, inflow Reynolds number, and inflow velocity profile (plug, contoured nozzle; and parabolic, straight tube). Normalized results exemplify a sensitive accurate means of validating, globally, reduced chemical kinetic models at approx. 1 atm and the relatively low temperatures approximating the loss of non-premixed idealized flameholding, e.g., in scramjet combustors. Laminar FS is defined locally as maximum air input velocity, U(sub air), that sustains combustion of a counter-jet of g-fuel at extinction. It uniquely characterizes a fuel. And global axial strain rate at extinction (U(sub air) normalized by nozzle or tube diameter, D(sub n or (sub t)) can be compared directly with computed extinction limits, determined using either a 1-D Navier Stokes stream-function solution, using detailed transport and finite rate chemistry, or (better yet) a detailed 2-D Navier Stokes numerical simulation. The experimental results define an idealized flameholding reactivity scale that shows wide ranging (50 x) normalized FS s for various vaporized-liquid and gaseous HCs, including, in ascending order: JP-10, methane, JP-7, n-heptane, n-butane, propane, ethane, and ethylene. Results from H2 air produce a unique and exceptionally strong flame that agree within approx. 1% of a recent 2-D numerically simulated FS for a 3 mm tube-OJB. Thus we suggest that experimental FS s and/or FS ratios, for various neat and blended HCs w/ and w/o additives, offer accurate global tests of chemical kinetic models at the Ts and Ps of extinction. In conclusion, we argue the FS approach is more direct and fundamental, for

  8. Slaughterhouse wastewater treatment: evaluation of a new three-phase separation system in a UASB reactor. (United States)

    Caixeta, Cláudia E T; Cammarota, Magali C; Xavier, Alcina M F


    The anaerobic treatment of the wastewater from the meat processing industry was studied using a 7.2 1 UASB reactor. The reactor was equipped with an unconventional configuration of the three-phase separation system. The effluent was characterized in terms of pH (6.3-6.6), chemical oxygen demand (COD) (2,000-6,000 mg l(-1)), biochemical oxygen demand BOD5 (1,300-2,300 mg 1(-1)), fats (40-600 mg l(-1)) and total suspended solids (TSS) (850-6,300 mg l(-1)) The reactor operated continuously throughout 80 days with hydraulic retention time of 14, 18 and 22 h. The wastewater from Rezende Industrial was collected after it had gone through pretreatment (screening, flotation and equalization). COD, BOD and TSS reductions and the biogas production rate were the parameters considered in analyzing the efficiency of the process. The average production of biogas was 111 day(-1) (STP) for the three experimental runs. COD removal varied from 77% to 91% while BOD removal was 95%. The removal of total suspended solids varied from 81% to 86%. This fact supports optimal efficiency of the proposed three-phase separation system as well as the possibility of applying it to the treatment of industrial effluents.

  9. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)


    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  10. Study of DNA damage with a new system for irradiation of samples in a nuclear reactor. (United States)

    Gual, Maritza R; Milian, Felix M; Deppman, Airton; Coelho, Paulo R P


    In this paper, we report results of a quantitative analysis of the effects of neutrons on DNA, and, specifically, the production of simple and double breaks of plasmid DNA in aqueous solutions with different concentrations of free-radical scavengers. The radiation damage to DNA was evaluated by electrophoresis through agarose gels. The neutron and gamma doses were measured separately with thermoluminescent detectors. In this work, we have also demonstrated usefulness of a new system for positioning and removing samples in channel BH#3 of the IEA-R1 reactor at the Instituto de Pesquisas Energéticas e Nucleares (Brazil) without necessity of interrupting the reactor operation. Copyright © 2010 Elsevier Ltd. All rights reserved.

  11. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system (United States)

    Christiansen, David W.; Smith, Bob G.


    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  12. A modular diagnosis system based on fuzzy logic for UASB reactors treating sewage. (United States)

    Borges, R M; Mattedi, A; Munaro, C J; Franci Gonçalves, R

    A modular diagnosis system (MDS), based on the framework of fuzzy logic, is proposed for upflow anaerobic sludge blanket (UASB) reactors treating sewage. In module 1, turbidity and rainfall information are used to estimate the influent organic content. In module 2, a dynamic fuzzy model is used to estimate the current biogas production from on-line measured variables, such as daily average temperature and the previous biogas flow rate, as well as the organic load. Finally, in module 3, all the information above and the residual value between the measured and estimated biogas production are used to provide diagnostic information about the operation status of the plant. The MDS was validated through its application to two pilot UASB reactors and the results showed that the tool can provide useful diagnoses to avoid plant failures.

  13. Passive containment cooling system with drywell pressure regulation for boiling water reactor (United States)

    Hill, Paul R.


    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  14. CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System

    DEFF Research Database (Denmark)

    Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil


    A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...

  15. Radiation safety assessment of a system of small reactors for distributed energy. (United States)

    Odano, N; Ishida, T


    A passively safe small reactor for a distributed energy system, PSRD, is an integral type of light-water reactor with a thermal output of 100 or 300 MW aimed to be used for supplying district heat, electricity to small grids, and so on. Candidate locations for the PSRD as a distributed energy source are on-ground, deep underground, and in a seaside pit in the vicinity of the energy consumption area. Assessments of the radiation safety of a PSRD were carried out for three cases corresponding to normal operation, shutdown and a hypothetical postulated accident for several siting candidates. Results of the radiation safety assessment indicate that the PSRD design has sufficient shielding performance and capability and that the exposure to the general public is very low in the case of a hypothetical accident.

  16. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao


    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  17. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.


    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  18. Duquesne Light Company`s burner modification for NO{sub x} RACT compliance on a 200 MW single face fired pulverized coal unit

    Energy Technology Data Exchange (ETDEWEB)

    Bionda, J.P. [Energy Systems Associates, Pittsburgh, PA (United States); Gabrielson, J.E.; Hallo, A.


    This paper discusses the result of a research test program conducted on Duquesne Light Company`s Elrama Unit 4. The program was designed to determine the viability of achieving compliance with the recently enacted PA DER Reasonably Available Control Technology (RACT) regulations. These regulations stipulate presumptive RACT requirements for wall fired boilers which include the installations and operation of low NO{sub x} burners with separated overfire air. Duquesne Light Company contracted Energy Systems, Associates (ESA) to aide in the design and testing of a novel low NO{sub x} burner design and separated overfire air system. By modifying the coal burners, it has been possible to reduce the NO{sub x} emissions by 50% to 60% on Unit 4, with minimal impact to the unburned carbon in the ash. The burner modifications create fuel rich streams which are surrounded by air rich zones in the primary flame region, thus staging combustion at the burner. Additional NO{sub x} reductions are realized when the combustion is further staged by use of the separated overfire air system.

  19. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.


    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  20. Attenuation capability of low activation-modified high manganese austenitic stainless steel for fusion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, M.M. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El-kameesy, S.U.; El-Fiki, S.A. [Physics Department, Faculty of Science, Ain Shams University, Cairo (Egypt); Ghali, S.N. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El Shazly, R.M. [Physics Department, Faculty of Science, Al-Azhar University, Cairo (Egypt); Saeed, Aly, E-mail: [Nuclear Power station Department, Faculty of Engineering, Egyptian-Russian University, Cairo (Egypt)


    Highlights: • Improvement stainless steel alloys to be used in fusion reactors. • Structural, mechanical, attenuation properties of investigated alloys were studied. • Good agreement between experimental and calculated results has been achieved. • The developed alloys could be considered as candidate materials for fusion reactors. - Abstract: Low nickel-high manganese austenitic stainless steel alloys, SSMn9Ni and SSMn10Ni, were developed to use as a shielding material in fusion reactor system. A standard austenitic stainless steel SS316L was prepared and studied as a reference sample. The microstructure properties of the present stainless steel alloys were investigated using Schaeffler diagram, optical microscopy, and X-ray diffraction pattern. Mainly, an austenite phase was observed for the prepared stainless steel alloys. Additionally, a small ferrite phase was observed in SS316L and SSMn10Ni samples. The mechanical properties of the prepared alloys were studied using Vickers hardness and tensile tests at room temperature. The studied manganese stainless steel alloys showed higher hardness, yield strength, and ultimate tensile strength than SS316L. On the other hand, the manganese stainless steel elongation had relatively lower values than the standard SS316L. The removal cross section for both slow and total slow (primary and those slowed down in sample) neutrons were carried out using {sup 241}Am-Be neutron source. Gamma ray attenuation parameters were carried out for different gamma ray energy lines which emitted from {sup 60}Co and {sup 232}Th radioactive sources. The developed manganese stainless steel alloys had a higher total slow removal cross section than SS316L. While the slow neutron and gamma rays were nearly the same for all studied stainless steel alloys. From the obtained results, the developed manganese stainless steel alloys could be considered as candidate materials for fusion reactor system with low activation based on the short life

  1. Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)


    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

  2. Instrumentation and control system for the prototype fast breeder reactor 'MONJU' power station

    Energy Technology Data Exchange (ETDEWEB)

    Hara, Hiroshi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)); Mae, Yoshinori; Ishida, Takayuki; Hashiura, Kazuhiko; Kasai, Shozo; Yamamoto, Hajime


    The fast breeder reactor 'Monju' power station is constructed as the nuclear power station of next generation in Tsuruga City, Fukui Prefecture. In order to realize high safety and operational reliability as the newest nuclear power station, the measurement and control system of Monju (electric power output 280 MW) has been designed and manufactured by reflecting the experiences of construction and operation of the experimental FBR 'Joyo' and the results of various research and development of sodium instrumentation and others, and by using the latest digital control technology and multiplexing system technology. In this paper, the results of development of the characteristic measurement and control technology as fast breeder reactors and the state of application to the measurement and control system which was designed and manufactured for Monju are described. Central monitoring panel, plant control system, sodium instrumentation, preheating control system and so on are reported. In the case of Monju, the heat capacity and thermal inertia of the primary and secondary cooling systems are large, and the system comprises three loops. (K.I.).

  3. Application of S-CO{sub 2} Cycle for Small Modular Reactor coupled with Desalination System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)


    The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO{sub 2} power cycle technology. The S-CO{sub 2} Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO{sub 2} cycles for SMART with desalination system is conducted. The simple recuperated S-CO{sub 2} cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%.

  4. Combustion of low calorific value gases in porous burners; Verbrennung von niederkalorischen Gasen in Porenbrennern

    Energy Technology Data Exchange (ETDEWEB)

    Diezinger, S.; Talukdar, P.; Issendorff, F. von; Trimis, D. [Lehrstuhl fuer Stroemungsmechanik Friedrich-Alexander-Univ., Erlangen-Nuernberg (Germany)


    By the use of low calorific value gases significant energy amounts can be saved, emissions can be reduced and system efficients can be increased. These mixtures are generated in different fields like waste sites and fuel cell systems with reformation of hydrocarbons. Conventional combustion techniques are not suited for the combustion of this kind of gases. Due to its high internal heat recuperation the porous burner technology has great potential for the combustion of low calorific value gases. In this work the influence of the combustion zone properties, the surface load and the educt temperature were determined by numerical simulations and experiments. (orig.)

  5. Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

    Energy Technology Data Exchange (ETDEWEB)

    Kunsman, David Marvin; Aldemir, Tunc (Ohio State University); Rutt, Benjamin (Ohio State University); Metzroth, Kyle (Ohio State University); Catalyurek, Umit (Ohio State University); Denning, Richard (Ohio State University); Hakobyan, Aram (Ohio State University); Dunagan, Sean C.


    This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other

  6. Numerical simulations of flow field in the target region of accelerator-driven subcritical reactor system

    CERN Document Server

    Chen Hai Yan


    Numerical simulations of flow field were performed by using the PHOENICS 3.2 code for the proposed spallation target of accelerator-driven subcritical reactor system (ADS). The fluid motion in the target is axisymmetric and is treated as a 2-D steady-state problem. A body-fitted coordinate system (BFC) is then chosen and a two-dimensional mesh of the flow channel is generated. Results are presented for the ADS target under both upward and downward flow, and for the target with diffuser plate installed below the window under downward flow

  7. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I


    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  8. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady


    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  9. Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program (United States)

    Ambrus, J. H.; Wright, W. E.; Bunch, D. F.


    The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

  10. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin


    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  11. Nitrifying-denitrifying filters and UV-C disinfection reactor: a combined system for wastewater treatment. (United States)

    Ben Rajeb, Asma; Mehri, Inès; Nasr, Houda; Najjari, Afef; Saidi, Neila; Hassen, Abdennaceur


    Biological treatment systems use the natural processes of ubiquitous organisms to remove pollutants and improve the water quality before discharge to the environment. In this paper, the nitrification/denitrification reactor allowed a reduction in organic load, but offered a weak efficiency in nitrate reduction. However, the additions of the activated sludge in the reactor improve this efficiency. A decrease of [Formula: see text] values from 13.3 to 8 mg/l was noted. Nevertheless, sludge inoculation led to a net increase of the number of pathogenic bacteria. For this reason, a UV-C pilot reactor was installed at the exit of the biological nitrification-denitrification device. Thus, a fluence of 50 was sufficient to achieve values of 20 MPN/100 ml for fecal coliform and 6 MPN/100 ml for fecal streptococci, conforms to Tunisian Standards of Rejection. On the other hand, the DGGE approach has allowed a direct assessment of the bacterial community changes upon the treated wastewater.

  12. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)


    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

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

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)


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

  14. Integrated photocatalytic and sequencing batch reactor (SBR) treatment system for degradation of phenol (United States)

    Yusoff, Nik Noor Athirah Nik; Ong, Soon-An; Ho, Li-Ngee; Wong, Yee-Shian; Khalik, Wan Fadhilah Wan Mohd; Lee, Sin-Li


    This study will examine the efficiency of the simultaneous photocatalytic and biodegradation process in the same treatment reactor. The sequencing batch reactor or also known as SBR is an effective wastewater treatment method that has been applied widely. SBR system has become an alternative method for industrial wastewater treatment with high concentration of chemical oxygen demand (COD), and phenolic compound. In order for the photocatalytic process to occur, ZnO nanoparticles immobilized onto sponge were introduced to the reactor. It was observed that the COD value were decreased, indicated that the simultaneous biodegradation and photodegradation process in functional. The effect of ZnO nanoparticles on the production and composition of extracellular polymeric substances (EPS) and the physiochemical stability of activated sludge in hybrid growth type SBR were monitored. The percentages of removal are varied with different concentration of ZnO nanoparticles. The highest COD removal recorded is 31.5% with concentration of ZnO 0.6 mg/L. With the present of the ZnO nanoparticles, the degradation of phenol was relatively better than combination of biological of photlysis and biological.

  15. Development of Integrated Regulatory Aging Management System related to Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Hanok; Park, Jeongsoon; Kim, Seonjae; Jhung, Myungjo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)


    The primary function of the reactor vessel internals (RVIs) is to support the core, the control rod assemblies, the core support structure and the reactor pressure vessel (RPV) surveillance capsules. The RVIs have the additional function to direct the flow of the reactor coolant and provide shielding for the RPV. Ageing mechanisms are specific processes that gradually change characteristics of a component with time and use. According to the Generic Aging Lessons Learned (GALL) report, aging mechanisms, such as fatigue, embrittlement, corrosion, wear, radiation induced creep, relaxation and swelling, is related to RVIs. Establishing that effects of aging degradation in RVIs are adequately managed is vital for assuring continued functionality of RVIs. To achieve this goal, it is necessary to develop the regulatory standard as well as generic inspection and evaluation guideline for RVIs. In this paper, the Integrated Regulatory Aging Management System (IR-Aging), which efficiently manages key data necessary to the development of regulatory standards and assists effective evaluation of RVIs, is proposed. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base. In this paper, the Integrated Regulatory Aging Management System (IR-Aging) is proposed in order to manage data necessary to the development of regulatory standards and assists effective evaluation of RVIs. The proposed system provides various documents, such as US NRC and domestic regulatory documents, licensee's documents submitted to a regulatory body, and research documents. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base.

  16. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas


    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  17. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux. (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A


    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  18. (I) A Declarative Framework for ERP Systems(II) Reactors: A Data-Driven Programming Model for Distributed Applications

    DEFF Research Database (Denmark)

    Stefansen, Christian Oskar Erik

    . • Using Soft Constraints to Guide Users in Flexible Business Process Management Systems. The paper shows how the inability of a process language to express soft constraints—constraints that can be violated occasionally, but are closely monitored—leads to a loss of intentional information in process....../Asynchronous Programming Model for Distributed Applications. The paper motivates, explains, and defines a distributed data-driven programming model. In the model a reactor is a stateful unit of distribution. A reactor specifies constructive, declarative constraints on its data and the data of other reactors in the style...

  19. Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)


    The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

  20. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim


    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  1. Experiments on a rotating-pipe swirl burner

    NARCIS (Netherlands)

    Hübner, A.W.; Tummers, M.J.; Hanjalic, K.; van der Meer, Theodorus H.


    Laser-Doppler measurements of mean velocity components and Reynolds stresses are reported in the near-field of a stable swirling flame of a newly designed natural-gas-fired experimental burner. The swirling motion is generated by the rotating outer pipe of the annular air passage, thus providing

  2. How Efficient is a Laboratory Burner in Heating Water? (United States)

    Jansen, Michael P.


    Describes an experiment in which chemistry students determine the efficiency of a laboratory burner used to heat water. The reaction is assumed to be the complete combustion of methane, CH4. The experiment is appropriate for secondary school chemistry students familiar with heats of reaction and simple calorimetry. Contains pre-laboratory and…

  3. Camping Burner-Based Flame Emission Spectrometer for Classroom Demonstrations (United States)

    Ne´el, Bastien; Crespo, Gasto´n A.; Perret, Didier; Cherubini, Thomas; Bakker, Eric


    A flame emission spectrometer was built in-house for the purpose of introducing this analytical technique to students at the high school level. The aqueous sample is sprayed through a homemade nebulizer into the air inlet of a consumer-grade propane camping burner. The resulting flame is analyzed by a commercial array spectrometer for the visible…

  4. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    Energy Technology Data Exchange (ETDEWEB)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)


    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

  5. Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

    Energy Technology Data Exchange (ETDEWEB)

    Bischi, Aldo


    Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

  6. Detailed Design of the Safety Residual Heat Removal System and a Circulation Pump for the KIJANG Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonghoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Primary cooling system (PCS) circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the fuel assembly in the reactor core is removed continuously. The PCS is designed based on the required thermal design flow rate of the reactor core, uncertainty of measuring instruments and the safe functions. Primary coolant is generally dumped into the pool and goes to the reactor core through the flow guide. The fission heat generated from the fuel assembly is transferred to the coolant, and then heated coolant goes to the PCS equipment room in order to remove the heat through the heat exchanger. SRHSR is designed based on the required flow rate and system constraints. Centrifugal pump of Case 1 with a non-dimensional specific speed of 0.97 [-] and specific diameter of 3.33 [-] is chosen as the SRHRS pump for the KJRR.


    Directory of Open Access Journals (Sweden)

    Anna Hajduk


    Full Text Available An alternative to aerobic wastewater treatment systems are anaerobic reactors. When designing anaerobic reactors attention is paid to the appropriate filling, pumping systems, or mixing systems, enabling the re-duction of technological limitations, which contribute to the improvement of end effects such as, quantity and quality of the resulting biogas and the quality of treated wastewater. Described experiment related to researches on the evaluation of the efficiency of removing contamina-tions from synthetic dairy waste water using anaerobic reactor equipped with an innovative mixing sys-tem. The efficiency of removal of organic compounds made studies ranged from 96.25% to 99.03%. The concentration of total nitrogen in raw wastewater was at a level of 148.36 ± 0 mg N/dm3 to 593.42 ± 94,92 mg N/dm3, and treated wastewater from 21.66 ± 19.71 mg N/dm3 to 28.73 ± 0.4 mg N/dm3. The concentration of total phosphorus in raw wastewater was at a level of 110 ± 0 mg P/dm3 to 441.16 ± 19.83 mg P/dm3, and treated wastewater from 16.49 ± 16.13 mg P/dm3 to 354 ± 14.18 mg P/dm3. The methane content of the biogas produced was at a level of from 0.0413 dm3 per 1 g COD introduced to 0.4367 dm3 per 1 g COD introduced.

  8. Elements of record management system for the RA research reactor decommissioning

    Directory of Open Access Journals (Sweden)

    Stejić Milijana


    Full Text Available According to latest recommendations, the record management system of a nuclear facility should operate as a part of the integrated management information system, and is implemented at the very beginning of the facility’s life cycle. The record management becomes particularly important at the end of the operation of a facility and then the operational record management system gradually transforms to a decommissioning one. However there is a significant number of nuclear facilities in the world which have reached the decommissioning stage with out having neither the initial decommissioning plan nor the established record management system. The objective of this paper is to introduce constituted elements of the record management system for the decommissioning of the RA research reactor in the VINČA Institute of Nuclear Sciences, and to discuss future planned actions related to this matter.

  9. Systems and methods for harvesting and storing materials produced in a nuclear reactor (United States)

    Heinold, Mark R.; Dayal, Yogeshwar; Brittingham, Martin W.


    Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

  10. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.


    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  11. Design and flow analysis for an oxygen-blown pulverized coal burner

    Energy Technology Data Exchange (ETDEWEB)

    Haeyang Pak; Nobuyuki Iwashima; Noriyuki Kobayashi; Masanobu Hasatani [Nagoya University, Nagoya (Japan). Department of Energy Engineering and Science


    An oxygen-blown pulverized coal burner for utilization of various kinds of coal was newly proposed and developed. The combustion efficiency of 99.7% was achieved by the moderate swirl burner. The flame stabilization could not be realized by the strong swirl burner, and the content of unburned carbon in ash was more than that of the moderate swirl burner experiment. The distribution of vorticity in the moderate swirling flow was equally proportioned even though the flow ratio was changed between 0.15 and 0.88. Additionally, the state of untidiness was observed near the central part of the burner nozzle in the strong swirling flow. The close relationship between combustion efficiency and vorticity profiles was found by PIV analysis of the flow. The moderate swirl burner was suitable for designing the burner structure in oxygen-blown pulverized coal combustion.

  12. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications (United States)

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


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

  13. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo


    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  14. Generation IV: new reactor systems; Neue Reaktorsysteme innerhalb der Generation IV Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). IKET; Hofmeister, J. [RWE Power AG, Regenerative Stromerzeugung, Essen (Germany); Tromm, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Programm Nukleare Sicherheitsforschung


    Generation IV, an initiative for international cooperation in nuclear technology, was launched by 10 states in 2000 and joined by Euratom in July 2003. Its aim is to assess nuclear energy systems complying with future safety, disposal, proliferation, and public acceptance requirements. The Forschungszentrum Karlsruhe focuses on design, thermohydraulics, and neutron kinetics. Work is mainly devoted to the high-performance light water reactor (HPLWR) with supercritical steam conditions. Thus, competence can be maintained, as the HPLWR issues qualify for later work in nuclear industry. (orig.)


    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick


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

  16. Afterheat removal from a helium reactor under accident conditions. CFD calculations for the code-to-code benchmark analyses on the thermal behavior for the gas turbine modular helium reactor

    Energy Technology Data Exchange (ETDEWEB)

    Siccama, N.B.; Koning, H


    The International Atomic Energy Agency (IAEA) Co-ordinated Research Programme (CRP) on `Heat Transport and Afterheat Removal for Gas Cooled Reactors under Accident Conditions` has organised benchmark analyses to support verification and validation of analytical tools used by the participants to predict the thermal behaviour of advanced gas cooled reactors during accidents. One of thew benchmark analyses concerns the code-to-code analysis of the Gas Turbine Modular Helium Reactor (GT-MHR) plutonium burner accidents. The GT-MHR is a passive safe, helium cooled, graphite moderated, advanced reactor system with a thermal power of 600 MW that is based on existing technology. The GT-MHR can also be fuelled with plutonium. If the main helium cooling and the auxiliary shut-down cooling systems fail or become unavailable, the core afterheat is removed by radiation and convection inside the reactor vessel and the reactor cavity to the Reactor Cavity Cooling System (RCCS). The objective of the RCCS is to serve as an ultimate heat sink, ensuring the thermal integrity of the core, vessel and critical equipment within the reactor cavity for the entire spectrum of postulated accident sequences. This paper describes the heat transport inside the reactor core to the RCCS. For this purpose, the heat transfer mechanisms as well as the flow patterns inside the core, the reactor pressure vessel, and the cavity have been calculated by the Computational Fluid Dynamics (CFD) code CFX-F3D. The behaviour of the RCCS itself is not described. One calculation considers the full power operation, while two calculations consider Loss Of Forced Convection (LOFC) accidents, one at pressurised conditions and the other depressurised conditions. The heat transfer from the reactor vessel to the environment under normal operation conditions is 2.64 MW. The highest temperature in the core is 1222K, and the average core temperature is 1075K. The highest reactor vessel temperature is 679K. The highest

  17. Impact of fuel quality and burner capacity on the performance of wood pellet stove

    Directory of Open Access Journals (Sweden)

    Petrović-Bećirović Sanja B.


    Full Text Available Pellet stoves may play an important role in Serbia in the future when fossil fuel fired conventional heating appliances are replaced by more efficient and environmentally friendly devices. Experimental investigation was conducted in order to examine the influence of wood pellet quality, as well as burner capacity (6, 8 and 10 kW, used in the same stove configuration, on the performance of pellet stove with declared nameplate capacity of 8 kW. The results obtained showed that in case of nominal load and combustion of pellets recommended by the stove manufacturer, stove efficiency of 80.03% was achieved. The use of lower quality pellet caused additional 1.13 kW reduction in heat output in case of nominal load and 0.63 kW in case of reduced load. This was attributed to less favourable properties and lower bulk and particle density of lower quality pellet. The use of different burner capacity has shown to have little effect on heat output and efficiency of the stove when pre-set values in the control system of the stove were not altered. It is concluded that replacement of the burner only is not sufficient to increase/decrease the declared capacity of the same stove configuration, meaning that additional measures are necessary. These measures include a new set up of the stove control system, which needs to be properly adjusted for each alteration in stove configuration. Without the adjustment mentioned, declared capacity of the stove cannot be altered, while its CO emission shall be considerably increased.

  18. ENFORM II: a calculational system for light water reactor logistics and effluent analysis

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Lewallen, M.A.; Purcell, W.L.; Cole, B.M.


    ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information.

  19. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources (United States)

    Juhasz, A. J.; Jones, B. I.


    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  20. Successional development of biofilms in moving bed biofilm reactor (MBBR) systems treating municipal wastewater. (United States)

    Biswas, Kristi; Taylor, Michael W; Turner, Susan J


    Biofilm-based technologies, such as moving bed biofilm reactor (MBBR) systems, are widely used to treat wastewater. Biofilm development is important for MBBR systems as much of the microbial biomass is retained within reactors as biofilm on suspended carriers. Little is known about this process of biofilm development and the microorganisms upon which MBBRs rely. We documented successional changes in microbial communities as biofilms established in two full-scale MBBR systems treating municipal wastewater over two seasons. 16S rRNA gene-targeted pyrosequencing and clone libraries were used to describe microbial communities. These data indicate a successional process that commences with the establishment of an aerobic community dominated by Gammaproteobacteria (up to 52 % of sequences). Over time, this community shifts towards dominance by putatively anaerobic organisms including Deltaproteobacteria and Clostridiales. Significant differences were observed between the two wastewater treatment plants (WWTPs), mostly due to a large number of sequences (up to 55 %) representing Epsilonproteobacteria (mostly Arcobacter) at one site. Archaea in young biofilms included several lineages of Euryarchaeota and Crenarchaeota. In contrast, the mature biofilm consisted entirely of Methanosarcinaceae (Euryarchaeota). This study provides new insights into the community structure of developing biofilms at full-scale WWTPs and provides the basis for optimizing MBBR start-up and operational parameters.

  1. Design and development of fast pneumatic transfer system (PTS) for instrumental neutron activation analysis at Jordan research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yongsam; Kim, Sunha; Moon, Jonghwa; Choi, Jinbok; Lee, Jongmin; Ryu, Jungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    A pneumatic transfer system (PTS) is one of the important equipment used for an neutron irradiation of a target material for an instrumental neutron activation analysis (INAA) in a research reactor. In particular, a rapid pneumatic transportation of irradiation capsule is essential for an accurate measurement of a short half-life nuclide. Three types of PTS for NAA facility at the Jordan Research and Training Reactor (JRTR) were newly developed for a functional improvement involving a manual and an automatic system which is equipped with programmable logic controller, software, and 13 devices to facilitate optimal operation of the system. In this paper, the designs and construction of these PTS, the operation and control of the system are described. In addition, a functional and operational test of the system were carried out as one of the basic requirement and characteristic parameters, and the results were reported to provide a user information as well as for the management and safety of the reactor.

  2. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau


    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  3. The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system (United States)

    Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid


    The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

  4. Low Emissions Burner Technology for Metal Processing Industry using Byproducts and Biomass Derived Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, Ajay; Taylor, Robert


    This research and development efforts produced low-emission burner technology capable of operating on natural gas as well as crude glycerin and/or fatty acids generated in biodiesel plants. The research was conducted in three stages (1) Concept definition leading to the design and development of a small laboratory scale burner, (2) Scale-up to prototype burner design and development, and (3) Technology demonstration with field vefiication. The burner design relies upon the Flow Blurring (FB) fuel injection based on aerodynamically creating two-phase flow near the injector exit. The fuel tube and discharge orifice both of inside diameter D are separated by gap H. For H < 0.25D, the atomizing air bubbles into liquid fuel to create a two-phase flow near the tip of the fuel tube. Pressurized two-phase fuel-air mixture exits through the discharge orifice, which results in expansion and breakup of air bubbles yielding a spray with fine droplets. First, low-emission combustion of diesel, biodiesel and straight VO (soybean oil) was achieved by utilizing FB injector to yield fine sprays for these fuels with significantly different physical properties. Visual images for these baseline experiments conducted with heat release rate (HRR) of about 8 kW illustrate clean blue flames indicating premixed combustion for all three fuels. Radial profiles of the product gas temperature at the combustor exit overlap each other signifying that the combustion efficiency is independent of the fuel. At the combustor exit, the NOx emissions are within the measurement uncertainties, while CO emissions are slightly higher for straight VO as compared to diesel and biodiesel. Considering the large variations in physical and chemical properties of fuels considered, the small differences observed in CO and NOx emissions show promise for fuel-flexible, clean combustion systems. FB injector has proven to be very effective in atomizing fuels with very different physical properties, and it offers a


    Energy Technology Data Exchange (ETDEWEB)



    This report discusses test campaign GCT4 of the Kellogg Brown & Root, Inc. (KBR) transport reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The transport reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using one of two possible particulate control devices (PCDs). The transport reactor was operated as a pressurized gasifier during GCT4. GCT4 was planned as a 250-hour test run to continue characterization of the transport reactor using a blend of several Powder River Basin (PRB) coals and Bucyrus limestone from Ohio. The primary test objectives were: Operational Stability--Characterize reactor loop and PCD operations with short-term tests by varying coal-feed rate, air/coal ratio, riser velocity, solids-circulation rate, system pressure, and air distribution. Secondary objectives included the following: Reactor Operations--Study the devolatilization and tar cracking effects from transient conditions during transition from start-up burner to coal. Evaluate the effect of process operations on heat release, heat transfer, and accelerated fuel particle heat-up rates. Study the effect of changes in reactor conditions on transient temperature profiles, pressure balance, and product gas composition. Effects of Reactor Conditions on Synthesis Gas Composition--Evaluate the effect of air distribution, steam/coal ratio, solids-circulation rate, and reactor temperature on CO/CO{sub 2} ratio, synthesis gas Lower Heating Value (LHV), carbon conversion, and cold and hot gas efficiencies. Research Triangle Institute (RTI) Direct Sulfur Recovery Process (DSRP) Testing--Provide syngas in support of the DSRP commissioning. Loop Seal Operations--Optimize loop seal operations and investigate increases to previously achieved maximum solids-circulation rate.

  6. Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors (United States)

    Lay, Wei-Siang

    Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled

  7. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.


    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  8. Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, W.P.; Bucher, R.G.


    Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

  9. Development of MCATHAS system of coupled neutronics/thermal-hydraulics in supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)


    The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)

  10. Bifurcation in the Lengyel–Epstein system for the coupled reactors with diffusion

    Directory of Open Access Journals (Sweden)

    Shaban Aly


    Full Text Available The main goal of this paper is to continue the investigations of the important system of Fengqi et al. (2008. The occurrence of Turing and Hopf bifurcations in small homogeneous arrays of two coupled reactors via diffusion-linked mass transfer which described by a system of ordinary differential equations is considered. I study the conditions of the existence as well as stability properties of the equilibrium solutions and derive the precise conditions on the parameters to show that the Hopf bifurcation occurs. Analytically I show that a diffusion driven instability occurs at a certain critical value, when the system undergoes a Turing bifurcation, patterns emerge. The spatially homogeneous equilibrium loses its stability and two new spatially non-constant stable equilibria emerge which are asymptotically stable. Numerically, at a certain critical value of diffusion the periodic solution gets destabilized and two new spatially nonconstant periodic solutions arise by Turing bifurcation.

  11. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee


    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  12. Development of Input/Output System for the Reactor Transient Analysis System (RETAS)

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik [ENESYS, Daejeon (Korea, Republic of); Ahn, Seung Hoon; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)


    A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view

  13. Requirement analysis and architecture of data communication system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, K. I.; Kwon, H. J.; Park, J. H.; Park, H. Y.; Koo, I. S


    When digitalizing the Instrumentation and Control(I and C) systems in Nuclear Power Plants(NPP), a communication network is required for exchanging the digitalized data between I and C equipments in a NPP. A requirements analysis and an analysis of design elements and techniques are required for the design of a communication network. Through the requirements analysis of the code and regulation documents such as NUREG/CR-6082, section 7.9 of NUREG 0800 , IEEE Standard 7-4.3.2 and IEEE Standard 603, the extracted requirements can be used as a design basis and design concept for a detailed design of a communication network in the I and C system of an integral reactor. Design elements and techniques such as a physical topology, protocol transmission media and interconnection device should be considered for designing a communication network. Each design element and technique should be analyzed and evaluated as a portion of the integrated communication network design. In this report, the basic design requirements related to the design of communication network are investigated by using the code and regulation documents and an analysis of the design elements and techniques is performed. Based on these investigation and analysis, the overall architecture including the safety communication network and the non-safety communication network is proposed for an integral reactor.

  14. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)


    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  15. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail:; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.


    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  16. Comprehensive Prediction of Thermosyphon Characteristics in Reactor Passive Cooling System Simulation Loop FASSIP-01

    Directory of Open Access Journals (Sweden)

    H. Tjahjono


    Full Text Available Passive cooling mechanism for a nuclear reactor has been proven to be very important since the Fukushima Daiichi Reactor accident that was caused by active cooling system malfunction due to total loss of electrical power source. In the Center for Nuclear Reactor Technology and Safety of BATAN, the cooling mechanism was studied by using a natural circulation test loop named FASSIP-01 that applied thermosyphon mechanism of water inside pipes of 1” diameter. This study aimed to analytically predictthe thermal characteristics of the loop including its response time towards steady condition usingthe MATLAB calculation program. This prediction derived the influence of several parameters such as the heat transfer coefficient of the cooler side (h-cooler, the heater power, the elevation difference between the heater and cooler(DZ, and the effects of the insulation thickness of pipe (IT on the flowrate, temperature, and the heat power distribution across all components in the loop. The result showed that byavoiding boiling condition, for transferring the heater power of 1000 W and 2000 W,the needed h-cooler exceeds 200 and 400 W m-2°C-1, respectively. For a h-cooler of 200 W m-2°C-1, the circulation flow rate increased from 0.04 to 0.06 kg/s-1 for heater power increase from 1000 W to 2000 W. Those flow rates were decreased to 0.037 and 0.052 kgs-1 by increasing h-cooler to 1000 W m-2°C-1.The results were in agreement with other studies on rectangular loops in the literature.The time needed to reach 95 % towards steady state was predicted to be more than 13 hours. Reduction of this time to less than five hours was possible by reducing the heater tank volume from 100 L to 30 L or by modifying the starting heater input power.

  17. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.


    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  18. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors (United States)

    Galvez, Cristhian


    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  19. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.


    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  20. Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)


    To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

  1. Co-firing straw with coal in a swirl-stabilized dual-feed burner: modelling and experimental validation. (United States)

    Yin, Chungen; Kaer, Søren K; Rosendahl, Lasse; Hvid, Søren L


    This paper presents a comprehensive computational fluid dynamics (CFD) modelling study of co-firing wheat straw with coal in a 150kW swirl-stabilized dual-feed burner flow reactor, in which the pulverized straw particles (mean diameter of 451microm) and coal particles (mean diameter of 110.4microm) are independently fed into the burner through two concentric injection tubes, i.e., the centre and annular tubes, respectively. Multiple simulations are performed, using three meshes, two global reaction mechanisms for homogeneous combustion, two turbulent combustion models, and two models for fuel particle conversion. It is found that for pulverized biomass particles of a few hundred microns in diameter the intra-particle heat and mass transfer is a secondary issue at most in their conversion, and the global four-step mechanism of Jones and Lindstedt may be better used in modelling volatiles combustion. The baseline CFD models show a good agreement with the measured maps of main species in the reactor. The straw particles, less affected by the swirling secondary air jet due to the large fuel/air jet momentum and large particle response time, travels in a nearly straight line and penetrate through the oxygen-lean core zone; whilst the coal particles are significantly affected by secondary air jet and swirled into the oxygen-rich outer radius with increased residence time (in average, 8.1s for coal particles vs. 5.2s for straw particles in the 3m high reactor). Therefore, a remarkable difference in the overall burnout of the two fuels is predicted: about 93% for coal char vs. 73% for straw char. As the conclusion, a reliable modelling methodology for pulverized biomass/coal co-firing and some useful co-firing design considerations are suggested. Copyright 2010 Elsevier Ltd. All rights reserved.

  2. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)


    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  3. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Reichenberger Michael A.


    Full Text Available Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional

  4. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.


    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  5. A complete miniaturized microstructured methanol fuel processor/fuel cell system for low power applications

    Energy Technology Data Exchange (ETDEWEB)

    Men, Yong; Kolb, Gunther; Zapf, Ralf; Tiemann, David; Wichert, Martin; Hessel, Volker; Loewe, Holger [Institut fuer Mikrotechnik Mainz GmbH, Carl Zeiss Str. 18-20, D-55129, Mainz (Germany)


    A complete miniaturized methanol fuel processor/fuel cell system was developed and put into operation as compact hydrogen supplier for low power application. The whole system consisting of a micro-structured evaporator, a micro-structured reformer and two stages of preferential oxidation of CO (PROX) reactor, micro-structured catalytic burner, and fuel cell was operated to evaluate the performance of the whole production line from methanol to electricity. The performance of micro methanol steam reformer and PROX reactor was systematically investigated. The effect of reaction temperature, steam to carbon ratio, and contact time on the methanol steam reformer performance is presented in terms of catalytic activity, selectivity, and reformate yield. The performance of PROX reactor fed with the reformate produced by the reformer reactor was evaluated by the variation of reaction temperature and oxygen to CO ratio. The results demonstrate that micro-structured device may be an attractive power source candidate for low power application. (author)

  6. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig


    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  7. Removal heat extraction systems in advanced reactors; Sistemas de extraccion de calor residual de la contencion en los reactores pasivos de tercera generacion

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J. C. de la; Munoz-Cobo, J. L.; Herranz, L.; Escriva, A.


    The two main problems generally attributed to the electricity generation by nuclear power are the security of the facility and the radioactivity of the nuclear wastes, in a way that the only tasks of the European Commission on this matter are to make sure a high level of security in the facilities, as well as an adequate fuel and waste management. In this paper we discuss about the main lines in which the CIEMAT and the Polytechnic University of Valencia are working relative to the study of the passive working systems of the advanced designs reactors. (Author) 24 refs.

  8. The zero age main sequence of WIMP burners (United States)

    Fairbairn, Malcolm; Scott, Pat; Edsjö, Joakim


    We modify a stellar structure code to estimate the effect upon the main sequence of the accretion of weakly-interacting dark matter onto stars and its subsequent annihilation. The effect upon the stars depends upon whether the energy generation rate from dark matter annihilation is large enough to shut off the nuclear burning in the star. Main sequence weakly-interacting massive particles (WIMP) burners look much like proto-stars moving on the Hayashi track, although they are in principle completely stable. We make some brief comments about where such stars could be found, how they might be observed and more detailed simulations which are currently in progress. Finally we comment on whether or not it is possible to link the paradoxically hot, young stars found at the galactic center with WIMP burners.

  9. Physicochemical properties of nanoparticles titania from alcohol burner calcination

    Directory of Open Access Journals (Sweden)

    Supan Yodyingyong


    Full Text Available The physicochemical properties of synthesized TiO2 nanoparticles from integrating sol-gel with flame-based techniques were studied. The synthesized nanoparticles properties were compared after using methanol, ethanol, and propanol fuel sources. The synthesized TiO2 were characterized by X-ray diffraction (XRD, transmission electron microscopy (TEM, thermal analysis (thermogravimetric analysis, TGA, and differential scanning calorimetry, DSC, and surface area Brunauer–Emmett–Teller (BET method. The photocatalytic activity of TiO2 nanoparticles was investigated by measuring the degradation of methylene blue. It was found that methanol and ethanol burners can be used as an alternative furnace that can yield TiO2 nanoparticles with physicochemical properties comparable to that of commercial TiO2 nanoparticles, while a propanol burner cannot be used as an alternative fuel.

  10. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.


    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.

  11. Discussion on polonium extraction systems for Pb-PI-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buongiorno, J. [Idaho National Engineering and Environmental Lab., Nuclear Engineering Dept., Idaho Falls, ID (United States); Larson, C.L.; Czerwinski, K.R. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering


    A discussion is presented on a polonium extraction technology that would reduce the radioactivity of the lead-bismuth coolant for fast reactors. This technology is based on the formation of the polonium hydride from the reaction of hydrogen gas with polonium-activated LBE. The equilibrium chemistry of the reaction was experimentally investigated. As a result, a correlation was generated for the free-energy of formation of the polonium hydride as a function of temperature. This correlation was then used for preliminary modeling of a polonium extraction system consisting in a mass exchanger where fine LBE droplets fall in countercurrent flow with a stream of pure hydrogen. It was found that a relatively compact and efficient polonium extraction system could be in principle designed, although significant technological and safety issues remain that are associated with the use and processing of hydrogen gas contaminated with polonium. (author)

  12. Research of lithium capillary-pore systems for fusion reactor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Evtikhin, V.A. E-mail:; Vertkov, A.V.; Lyublinski, I.E.; Khripunov, B.I.; Petrov, V.B.; Mirnov, S.V


    To date there is no adequate solution for high heat load plasma facing components of the next step fusion reactor among solid material options. A lithium-filled capillary porous systems (CPS) was proposed as a plasma facing material and experimental work on this subject is now in progress. Steady-state experiments with CPS-based target and lithium supply systems have shown successful operation at heat fluxes of 1-10 MW/m{sup 2} during several hours. Experimental data is obtained on lithium CPS stability at heat flux up to 25-50 MW/m{sup 2}. The lithium CPS behaviour in contact with real tokamak plasma is considered for normal discharge condition at 10 MW/m{sup 2} and for plasma disruption at 15 MJ/m{sup 2}. Erosion mechanism of lithium under tokamak plasma impact was analysed. Stability of lithium CPS in tokamak conditions was shown.

  13. 3-D Monte Carlo analyses of the shielding system in a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gallina, M.; Petrizzi, L.; Rado, V. (ENEA, Frascati (Italy). Centro Ricerche Energia)


    As part of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the shielding system performance in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) transport code (3-B version). The main issue is the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. ''Self generated weight windows'' and source biasing technique have been used to treat deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted. (author).

  14. 3-D Monte Carlo analyses of shielding system in tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gallina, M.; Petrizzi, L.; Rado, V.


    Within the framework of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations were carried out to assess system shielding performances in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) code (3-B version). The main issue concerns the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. 'Self generated weight windows' (w.w.) and source biasing techniques were used to treat the deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted.

  15. Method for carrying out biotechnological processes by means of a multi-phase system in a loop reactor.

    NARCIS (Netherlands)

    Tramper, J.


    A method for carrying out biotechnological processes by means of multiphase system in a loop reactor, which system comprises an aqueous phase (11) and at least one organic solvent (9, 10) which is immiscible with water and which has a different density from water; one of the liquid components is

  16. The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmal, N.; Leder, H.J.; Schade, H.J.


    Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated.

  17. Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system (United States)

    Christiansen, David W.; Smith, Bob G.


    A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

  18. Energy recovery from dairy waste-waters: impacts of biofilm support systems on anaerobic CST reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ramasamy, E.V.; Abbasi, S.A. [Pondicherry Central Univ., Centre for Pollution Control and Energy Technology, Pondicherry (India)


    Anaerobic digestion is one of the major steps involved in the treatment of dairy industry waste-waters and many CSTRs (continuously-stirred tank reactors) are functioning for this purpose all over the world. In this paper, the authors describe their attempts to upgrade a CSTR's performance by incorporating a biofilm support system (BSS) within the existing reactor. The focus of the work was to find an inexpensive and easy to install BSS which could significantly enhance the rates of waste treatment and methane recovery. Rolls of nylon mesh (with {approx}1 mm openings), of 5 cm height and 2 cm dia, when incorporated in the CSTR at the biofilm surface (with a digester volume ratio 0.3 cm{sup 2}/cm{sup 3}), enabled the CSTR to perform better with > 20% improvement in the methane yield. Such simple BSS devices can significantly improve the performance of a CSTR anaerobic digester treating dairy wastes. The enhancement is due to the development of active biofilms which not only enhance the micro-organism-waste contact but also reduce the microbial washout. Such devices are inexpensive and very easy to incorporate -- the gains are thus achieved with very little cost and effort. (Author)

  19. A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station (United States)

    Bloomfield, H. S.; Heller, J. A.


    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

  20. Structure of diffusion flames from a vertical burner (United States)

    Mark A. Finney; Dan Jimenez; Jack D. Cohen; Isaac C. Grenfell; Cyle Wold


    Non-steady and turbulent flames are commonly observed to produce flame contacts with adjacent fuels during fire spread in a wide range of fuel bed depths. A stationary gas-fired burner (flame wall) was developed to begin study of flame edge variability along an analagous vertical fuel source. This flame wall is surrogate for a combustion interface at the edge of a deep...

  1. Effect of cycled combustion ageing on a cordierite burner plate

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Eugenio [Instituto de Ceramica y Vidrio, CSIC, c/ Kelsen 5, Campus de Cantoblanco, 28049 Madrid (Spain); Gancedo, J. Ramon [Instituto de Quimica Fisica ' Rocasolano' , CSIC, c/ Serrano 119, 28006 Madrid (Spain); Gracia, Mercedes, E-mail: [Instituto de Quimica Fisica ' Rocasolano' , CSIC, c/ Serrano 119, 28006 Madrid (Spain)


    A combination of {sup 57}Fe-Moessbauer spectroscopy and X-ray Powder Diffraction analysis has been employed to study modifications in chemical and mechanical stability occurring in a cordierite burner aged under combustion conditions which simulate the working of domestic boilers. Moessbauer study shows that Fe is distributed into the structural sites of the cordierite lattice as Fe{sup 2+} and Fe{sup 3+} ions located mostly at octahedral sites. Ferric oxide impurities, mainly hematite, are also present in the starting cordierite material accounting for {approx_equal}40% of the total iron phases. From Moessbauer and X-ray diffraction data it can be deduced that, under the combustion conditions used, new crystalline phases were formed, some of the substitutional Fe{sup 3+} ions existing in the cordierite lattice were reduced to Fe{sup 2+}, and ferric oxides underwent a sintering process which results in hematite with higher particle size. All these findings were detected in the burner zone located in the proximity of the flame and were related to possible chemical reactions which might explain the observed deterioration of the burner material. Research Highlights: {yields}Depth profile analyses used as a probe to understand changes in refractory structure. {yields}All changes take place in the uppermost surface of the burner, close to the flame. {yields}Reduction to Fe{sup 2+} of substitutional Fe{sup 3+} ions and partial cordierite decomposition. {yields}Heating-cooling cycling induces a sintering of the existing iron oxide particles. {yields}Chemical changes can explain the alterations observed in the material microstructure.

  2. Assessment of the Implementation of a Neutron Measurement System During the Commissioning of the Jordan Research and Training Reactor

    Directory of Open Access Journals (Sweden)

    Sanghoon Bae


    Full Text Available The Jordan Research and Training Reactor (JRTR is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

  3. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail:


    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  4. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    Energy Technology Data Exchange (ETDEWEB)

    Gallup, D.R.


    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  5. Designing visual displays and system models for safe reactor operations based on the user`s perspective of the system

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.


    Most designers are not schooled in the area of human-interaction psychology and therefore tend to rely on the traditional ergonomic aspects of human factors when designing complex human-interactive workstations related to reactor operations. They do not take into account the differences in user information processing behavior and how these behaviors may affect individual and team performance when accessing visual displays or utilizing system models in process and control room areas. Unfortunately, by ignoring the importance of the integration of the user interface at the information process level, the result can be sub-optimization and inherently error- and failure-prone systems. Therefore, to minimize or eliminate failures in human-interactive systems, it is essential that the designers understand how each user`s processing characteristics affects how the user gathers information, and how the user communicates the information to the designer and other users. A different type of approach in achieving this understanding is Neuro Linguistic Programming (NLP). The material presented in this paper is based on two studies involving the design of visual displays, NLP, and the user`s perspective model of a reactor system. The studies involve the methodology known as NLP, and its use in expanding design choices from the user`s ``model of the world,`` in the areas of virtual reality, workstation design, team structure, decision and learning style patterns, safety operations, pattern recognition, and much, much more.

  6. Combustion Characteristics of Butane Porous Burner for Thermoelectric Power Generation

    Directory of Open Access Journals (Sweden)

    K. F. Mustafa


    Full Text Available The present study explores the utilization of a porous burner for thermoelectric power generation. The porous burner was tested with butane gas using two sets of configurations: single layer porcelain and a stacked-up double layer alumina and porcelain. Six PbSnTe thermoelectric (TE modules with a total area of 54 cm2 were attached to the wall of the burner. Fins were also added to the cold side of the TE modules. Fuel-air equivalence ratio was varied between the blowoff and flashback limit and the corresponding temperature, current-voltage, and emissions were recorded. The stacked-up double layer negatively affected the combustion efficiency at an equivalence ratio of 0.20 to 0.42, but single layer porcelain shows diminishing trend in the equivalence ratio of 0.60 to 0.90. The surface temperature of a stacked-up porous media is considerably higher than the single layer. Carbon monoxide emission is independent for both porous media configurations, but moderate reduction was recorded for single layer porcelain at lean fuel-air equivalence ratio. Nitrogen oxides is insensitive in the lean fuel-air equivalence ratio for both configurations, even though slight reduction was observed in the rich region for single layer porcelain. Power output was found to be highly dependent on the temperature gradient.

  7. NOx Pollution Analysis for a Sulfur Recovery Unit Thermal Reactor (United States)

    Yeh, Chun-Lang


    A sulfur recovery unit (SRU) thermal reactor is the most important equipment in a sulfur plant. It is negatively affected by high temperature operations. In this paper, NOx emissions from the SRU thermal reactors are simulated. Both the prototype thermal reactor and its modifications, including changing fuel mass fraction, changing inlet air quantity, changing inlet oxygen mole fraction, and changing burner geometry, are analyzed to investigate their influences on NOx emissions. In respect of the fuel mass fraction, the simulation results show that the highest NO emission occurs at a zone 1 fuel mass fraction of 0.375, around which the reactor maximum temperature and the zone 1 average temperature reach maximum values. Concerning the inlet air quantity, the highest NO emission occurs when the inlet air quantity is 2.4 times the designed inlet air quantity. This is very close to the inlet air quantity at which the maximum average temperature occurs. Regarding the inlet oxygen mole fraction, the NO emission increases as the inlet oxygen mole fraction increases. With regard to the burner geometry, the NO emission increases as the clearance of the burner acid gas tip increases. In addition, the NO emission increases as the swirling strength increases.

  8. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)


    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

  9. Instrumentation Needs for Integral Primary System Reactors (IPSRs) - Task 1 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Gary D. Storrick; Bojan Petrovic; Luca Oriani; Lawrence E. Conway; Diego Conti


    This report presents the results of the Westinghouse work performed under Task 1 of this Financial Assistance Award and satisfies a Level 2 Milestone for the project. While most of the signals required for control of IPSRs are typical of other PWRs, the integral configuration poses some new challenges in the design or deployment of the sensors/instrumentation and, in some cases, requires completely new approaches. In response to this consideration, the overall objective of Task 1 was to establish the instrumentation needs for integral reactors, provide a review of the existing solutions where available, and, identify research and development needs to be addressed to enable successful deployment of IPSRs. The starting point for this study was to review and synthesize general characteristics of integral reactors, and then to focus on a specific design. Due to the maturity of its design and availability of design information to Westinghouse, IRIS (International Reactor Innovative and Secure) was selected for this purpose. The report is organized as follows. Section 1 is an overview. Section 2 provides background information on several representative IPSRs, including IRIS. A review of the IRIS safety features and its protection and control systems is used as a mechanism to ensure that all critical safety-related instrumentation needs are addressed in this study. Additionally, IRIS systems are compared against those of current advanced PWRs. The scope of this study is then limited to those systems where differences exist, since, otherwise, the current technology already provides an acceptable solution. Section 3 provides a detailed discussion on instrumentation needs for the representative IPSR (IRIS) with detailed qualitative and quantitative requirements summarized in the exhaustive table included as Appendix A. Section 3 also provides an evaluation of the current technology and the instrumentation used for measurement of required parameters in current PWRs. Section 4

  10. The Programmable Logic Controller and its application in nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Palomar, J.; Wyman, R. [Lawrence Livermore National Lab., CA (United States)


    This document provides recommendations to guide reviewers in the application of Programmable Logic Controllers (PLCS) to the control, monitoring and protection of nuclear reactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the protection system engineering organization. Supplementing this document are two appendices. Appendix A summarizes PLC characteristics. Specifically addressed are those characteristics that make the PLC more suitable for emergency shutdown systems than other electrical/electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create an unsafe operating environment. Appendix B provides an overview of the use of programmable logic controllers in emergency shutdown systems. The intent is to familiarize the reader with the design, development, test, and maintenance phases of applying a PLC to an ESD system. Each phase is described in detail and information pertinent to the application of a PLC is pointed out.

  11. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    R. Ferrer; M. Asgari; S. Bays; B. Forget


    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

  12. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor; Desarrollo de la interface de usuario para la visualizacion de los sistemas auxiliares del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Merced D, J. E.


    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  13. Automatic control system in the reactor peggy; Systeme de pilotage automatique du reacteur peggy

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, J.; Mourchon, R.; Da Costa, D.; Desandre-Navarre, Ch. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires


    The equipment makes it possible for the reactor to attain a given power automatically and for the power to be maintained around this level. The principle of its operation consists in the changing from one power to another, at constant period, by means of a programmer transforming a power-step request into a voltage variation which is linear with time and which represents the logarithm of the required power. The real power is compared continuously with the required power. Stabilization occurs automatically as soon as the difference between the reactor power and the required power diminishes to a few per cent. (authors) [French] L'equipement permet au reacteur d'atteindre automatiquement une puissance affichee et de le stabiliser autour de cette puissance. Le principe du fonctionnement consiste a passer, a periode constante, d'une puissance a une autre, grace a un programmeur transformant une demande de puissance-echelon en une variation de tension lineaire en fonction du temps, tension representant le logarithme de la puissance affichee. La puissance reelle est comparee en permanence a la puissance affichee. La stabilisation intervient automatiquement lorsque la puissance du reacteur ne differe plus que de quelques pour cent de la valeur affichee. (auteurs)

  14. 10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

    Energy Technology Data Exchange (ETDEWEB)


    This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

  15. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Son, Han Seong; Song, Deok Yong [ENESYS, Taejon (Korea, Republic of); Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)


    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors.

  16. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)


    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  17. Direct ignition of pulverized coal. A new burner for a 600 MWe boiler

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J.P.; Malaubier, F.; Mevel, J.C.

    Operation of pulverized coal boilers requires gas or oil burners for starting up which is expensive especially in France where electricity is produced by nuclear energy and for peak hours by coal burners. A 1 MW pilot plant was developed in 1983 and an industrial 10.7 MW burner for a 600 MW boiler was built in 1985 and tested in 1986. Results are reported.

  18. Gas fired boilers: Perspective for near future fuel composition and impact on burner design process (United States)

    Schiro, Fabio; Stoppato, Anna; Benato, Alberto


    The advancements on gas boiler technology run in parallel with the growth of renewable energy production. The renewable production will impact on the fuel gas quality, since the gas grid will face an increasing injection of alternative fuels (biogas, biomethane, hydrogen). Biogas allows producing energy with a lower CO2 impact; hydrogen production by electrolysis can mitigate the issues related to the mismatch between energy production by renewable and energy request. These technologies will contribute to achieve the renewable production targets, but the impact on whole fuel gas production-to-consumption chain must be evaluated. In the first part of this study, the Authors present the future scenario of the grid gas composition and the implications on gas fed appliances. Given that the widely used premixed burners are currently designe