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Sample records for burner test reactor

  1. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    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

    2008-12-16

    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

  2. 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

    2006-09-15

    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.

  3. High conversion burner type reactor

    International Nuclear Information System (INIS)

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

  4. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  5. 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)

    2015-12-21

    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.

  6. Advanced Burner Reactor Preliminary NEPA Data Study

    International Nuclear Information System (INIS)

    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

  7. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    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

    2007-10-15

    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

  8. Tests of gas-blast burners

    International Nuclear Information System (INIS)

    Testing of the most sold small gas-blast burners on the Danish market was carried out with regard to carbon monoxide emission contra the content of oxygen in the flue gas in relation to the burners' combustion stability at varying fire box pressures. The burners tested were Weishaupt WG 1: DG no. 2506, Riello 40 GS3: DG no. 2722, Bentone BEG 15: DG no. 2153 and Box 1 G: no. 1104. This covers 90% of the Danish market for gas burners. It was concluded that all the burners had a broader area of adjustment possibilities without carbon monoxide emission than previously tested box burners. This with the exception of when surplus oxygen is low, where large of amounts of carbon monoxide are generated at an oxygen content in flue gas of ca. 2% (10.8% CO2). Burners in which the total pressure in the blower was high were the most stable with regard to air supply and varying fire-box pressure. It is pointed out that other conditions of design have also influence in this respect. In the cases of Weishaupt, Bentone and Riello burners there is a significant relation between blast pressure and oxygen content in the flue gas, whereas in the case of the Box burner, the percentage of oxygen in the flue gas rises in relation to increased pressure in the smoke outlet. The results of the tests are presented in great detail. (AB)

  9. 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

    2007-06-30

    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

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

    International Nuclear Information System (INIS)

    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 from normal high

  11. Ball plasma dynamics for FBX BURNER reactor

    International Nuclear Information System (INIS)

    The authors have been conducting fundamental experiments on the moving plasma balls in the major axis direction. This has a fundamental importance for the quasi-steady fusion reactor scheme FBX BURNER. This configuration is the descendants of Spheromak type nuclear fusion scheme but with long toroidal field coils. The main issues have been focused onto the dynamic stability of the moving plasma balls. The second issue is the collision between two successive plasma balls as shown. This shows a basic result from an optical measurement. The comparison with magnetic measurements shows peculiar but interesting features of the system. The dimension of the plasma injector is 2 m in length as a whole. It is in a coaxial configuration. The external electrode is 150 mm in inner diameter and the inner electrode is 50 mm in diameter. This is attached to a 1 m insulation reservoir with same inner diameter. An axial magnetic field up to 0.1 Tesla is applied on the discharge with a current of up to 10 kA in few ms. The authors show their experimental and numerical simulation results on these problems

  12. Linear accelerator for burner-reactor

    International Nuclear Information System (INIS)

    Future development of nuclear power engineering depends on the successful solution of two key problems of safety and utilization of high level radioactive wastes (HLRW) of atomic power plants (APP). Modern methods of HLRW treatment involve solidification, preliminary storing for a period of 30-50 years necessary for the decay of long-living nuclides and final burial in geological formations several hundred meters below the ground surface. The depth burial of the radioactive wastes requires complicated under ground constructions. It's very expensive and doesn't meet modern ecological requirements. Alternative modern and more reasonable methods of APP HLRW treatment are under consideration now. One of the methods involves separation of APP waste radionuclides for use in economy with subsequent transmutation of the long-living isotopes into the short-living ones by high-intensity neutron fluxes generated by proton accelerators. The installation intended for the long-living radionuclides transmutation into the short-living ones is called burner-reactor. It can be based on the continuous regime proton accelerator with 1.5 GeV energy, 0.3 A current and beam mean power of 450 MW. The preferable type of the proton accelerator with the aforementioned parameters is the linear accelerator

  13. 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

    2007-06-30

    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

  14. Carbide and Nitride Fuels for Advanced Burner Reactor

    International Nuclear Information System (INIS)

    The impacts of the mixed carbide and nitride fuels on the core performances and passive safety features of TRU burner were assessed and comapred with the metallic and oxide fuels. Targeting the potential design goals adopted in the Advanced Burner Reactor core concepts, the alternative TRU burner concepts were developed by loading carbide and nitride fuels. The neutron spectrum is softer than that of the metal core, but harder than that of the oxide core, and the core performance parameters such as fuel residence time, discharge burnup, flux level, etc are generally between the values of the metal and oxide cores. The margin to fuel melt was significantly increased because of the high thermal conductivity and high melting temperature, and hence there is an additional room to improve the thermal efficiency by increasing the operating temperature. The changed fuel composition affected the kinetics parameters and reactivity feedback coefficients, but the variations were minimal. The reduced core height decreases the sodium void worth, and the high thermal conductivity decreases the fuel temperature and Doppler constant. As a result, both carbide and nitride cores have favorable passive safety features without additional design fixes that are required in the oxide core concepts. (author)

  15. Investigation of CANDU reactors as a thorium burner

    International Nuclear Information System (INIS)

    Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in the form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates the prospects of exploiting the rich world thorium reserves in CANDU reactors. Two different fuel compositions have been selected for investigations: (1) 96% thoria (ThO2) + 4% PuO2 and (2) 91% ThO2 + 5% UO2 + 4% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the reactor criticality k ∞ and the burn-up values of the reactor have been pursued by full power operation for >∼8 years. The reactor starts with k ∞ = ∼1.39 and decreases asymptotically to values of k ∞ > 1.06, which is still tolerable and useable in a CANDU reactor. The reactor criticality k ∞ remains nearly constant between the 4th year and the 7th year of plant operation, and then, a slight increase is observed thereafter, along with a continuous depletion of the thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn-up can be achieved with the same fuel (>160,000 MW D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burned up, provided that the fuel rods could be fabricated to withstand such high burn-up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically

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

    International Nuclear Information System (INIS)

    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

  17. Fast Burner Reactor Devoted to Minor Actinide Incineration

    International Nuclear Information System (INIS)

    This study proposes a new fast reactor core concept dedicated to plutonium and minor actinide burning by transmutation. This core has a large power level of ∼1500 MW(electric) favoring the economic aspect. To promote plutonium and minor actinide burning as much as possible, total suppression of 238U, which produces 239Pu by conversion, and large quantities of minor actinides in the core are desirable. Therefore, the 238U-free fuel is homogeneously mixed with a considerable quantity of minor actinides.From the safety point of view, both the Doppler effect and the coolant (sodium) void reactivity become less favorable in a 238U-free core. To preserve these two important safety parameters on an acceptable level, a hydrogenated moderator separated from the fuel and nuclides, such as W or 99Tc, is added to the core in the place of 238U. Tungsten and 99Tc have strong capture resonances at appropriate energies, and 99Tc itself is a long-lived fission product to be transmuted with profit.This core allows the achievement of a consumption rate of ∼100 kg/TW(electric).h of transuranic elements, ∼70 kg/TW(electric).h for plutonium (due to 238U suppression), and 30 to 35 kg/TW(electric).h for minor actinides. In addition, ∼14 kg/TW(electric).h of 99Tc is destroyed when this element is present in the core (the initial loading of 99Tc is >4000 kg in the core).The activity of newly designed subassemblies has also been investigated in comparison to standard fast reactor subassemblies (neutron sources, decay heat, and gamma dose rate). Finally, a transmutation scenario involving pressurized water reactors and minor actinide-burning fast reactors has been studied to estimate the necessary proportion of burner reactors and the achievable radiotoxicity reduction with respect to a reference open cycle

  18. 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

    2008-02-01

    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.

  19. A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

    International Nuclear Information System (INIS)

    Research highlights: → This article discusses use of WG-plutonium as the startup fuel for Advanced Burner Reactor. → The presence of gallium in WG fuel may compromise the fuel integrity. → There is no facility exists to remove gallium from plutonium except at laboratory scale. → This article discusses the processes and issues associated with the gallium removal. → The article provides realistic scenario to all stack-holders involved in designing and operating ABR. - Abstract: The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the

  20. Demonstration test of burner liner strain measuring system

    Science.gov (United States)

    Stetson, K. A.

    1984-01-01

    A demonstration test was conducted for two systems of static strain measurement that had been shown to have potential for application jet engine combustors. A modified JT12D combustor was operated in a jet burner test stand while subjected simultaneously to both systems of instrumentation, i.e., Kanthal A-1 wire strain gages and laser speckle photography. A section of the burner was removed for installation and calibration of the wire gages, and welded back into the burner. The burner test rig was modified to provide a viewing port for the laser speckle photography such that the instrumented section could be observed during operation. Six out of ten wire gages survived testing and showed excellent repeatability. The extensive precalibration procedures were shown to be effective in compensating for the large apparent strains associated with these gages. Although all portions of the speckle photography system operated satisfactorily, a problem was encountered in the form of optical inhomogeneities in the hot, high-pressure gas flowing by the combustor case which generate large and random apparent strain distributions.

  1. The energy analysis of burner reactor power systems

    International Nuclear Information System (INIS)

    Currently most commercial nuclear power stations are based on thermal reactor designs called burner reactors which are net consumers of fissile material. These power stations form one part of a larger system that generates electricity from uranium. However, in addition to producing energy, such systems also consume energy, in the form of various fuels, during construction and operation. This thesis describes the use of energy analysis to determine the total energy required by these systems. A number of factors are shown to influence energy consumption and, in particular, the effect of extracting uranium from different sources is studied in detail. For ores, an important inverse relationship between energy use and ore grade is investigated and quantified. The physical limit at which the energy input to the system is equal to its output is shown to correspond to an average grade of 15 parts per million of 'triuranium octoxide'. Analysis of proposals for extracting uranium from seawater indicates that the only schemes giving a positive energy balance are costly (500 dollars/lb U3O8) and limited to low production rates. The effects of feedback within fuel systems are analysed and the results are used to formulate an economic model in which nuclear electricity prices determine uranium ore costs as well as vice versa. The model demonstrates that, with present techniques, the average economic limit to ore grade is 50 ppm U3O8 with subsequent resources, on current assessment, of only 107 tonnes U3O8. This contradicts most traditional studies which, by assuming fixed, non-dependent fuel costs, suggest an ore grade limit of less than 4 ppm U3O8 and economically recoverable resources in excess of 1010 tonnes U3O8. (author)

  2. Advanced fuel cycle scenario study in the European context using different burner reactor concepts

    International Nuclear Information System (INIS)

    Different types of fast spectrum dedicated burners have been proposed for the management of radioactive wastes in the frame of various advanced fuel cycle scenarios. Accelerator-driven systems (ADS) and critical low conversion ratio fast reactors have been studied, e.g. within the European context. A potential alternative system is a fusion-fission hybrid (FFH). In the present study, a sodium-cooled fast reactor driven by a D-T fusion neutron source, the subcritical advanced burner reactor (SABR) system is considered. In order to intercompare the different systems, a systematic study is under way. The performances of the two types of systems (SABR, ADS) will be compared from a minor actinide (MA) or transuranic (i.e. Pu+MA) burning potential point of view. The present paper reports preliminary results of the first phase of study, i.e. the comparison of SABR and ADS when used as minor actinides burners. (authors)

  3. Neutron economy and transmutation performance of coupling system of fast reactor and a-burner

    International Nuclear Information System (INIS)

    Neutron economy and transmutation performance are examined for a fast reactor (FR), a PWR, and an A-Burner which transmutes Minor-Actinide (MA) in a well-thermalized neutron field and a slightly hard neutron field optimized for the burn-up of 246Cm. The neutron economies of the FR and the A-Burner are, respectively, favorable and acceptable to transmute MA. The coupling system of both the reactors can reduce 8.1 ton of MA to almost zero within 60 years by using one FR and one A-Burner. This coupling system is expected to achieve the final goal of transmutation, i.e., to make us free from the geological disposal. (authors)

  4. Carbide and nitride fuels for advanced burner reactor

    International Nuclear Information System (INIS)

    Full text: Under the U.S. fast reactor program, reference and alternative 1000 MWth Advanced Burner Reactor (ABR) core concepts were developed using ternary metallic (U-TRU-Zr) and mixed oxide (UO2+TRUO2) fuels. Recently, mixed carbide and nitride fuels have been considered as fast reactor fuels on the basis of their high density, compatibility with coolant, high melting temperature, and excellent thermal conductivity although they are ceramic fuel like a mixed oxide fuel. Thus, the performance of the ABR core loaded with carbide and nitride fuels was evaluated in this study with an expectation that the carbide and nitride fuels can mitigate disadvantages of both metallic and oxide fuels in the ABR: favorable passive safety features in a severe accident compared to the oxide core, a higher discharge burnup compared to the metallic core, and a potential to increase thermal efficiency. All calculations performed in this study were focused on the neutronics characteristics, although the fabrication and irradiation experiences for carbide and nitride fuels are limited and some problems were observed in the reprocessing and irradiation of these fuels. The mixed monocarbide and mixed mononitride fuels were selected as the alternative fuel forms and the ABR core concepts with these fuels were developed based on the reference 1000 MWth ABR core concepts. For consistency, the potential design goals used in the reference ABR core concepts were also employed in this study: a 1000 MWth power rating, medium TRU conversion ratio of ∼0.75, a compact core, one-year operational cycle length at least with a capacity factor of 90%, sufficient shutdown margin with a limited maximum single control assembly fault, and possible use of either metallic or any ceramic fuels in the same core layout. The core layout and outer assembly dimensions of the reference 1000 MWth ABR core were kept, but the intra assembly design parameters were varied to maximize the discharge burnup within the

  5. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  6. ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor

    International Nuclear Information System (INIS)

    Description of program or function: In order to intercompare the characteristics of the different reactors considered for Pu recycling, in terms of neutron economy, minor actinide production, uranium content versus Pu burning, the NSC Working Party on Physics of Plutonium Recycling (WPPR) is setting up several benchmark studies. They cover in particular the case of the evolution of the Pu quality and Pu fissile content for Pu recycling in PWRs; the void coefficient in PWRs partly fuelled with MOX versus Pu content; the physics characteristics of non-standard fast reactors with breeding ratios around 0.5. The following benchmarks are considered here: - Fast reactors: Pu Burner MOX fuel, Pu Burner metal fuel; - PWRs: MOX recycling (bad quality Pu), Multiple MOX recycling

  7. Fuel cycle of actinide burner-reactor. Review of investigations by > program

    International Nuclear Information System (INIS)

    The problem of long-lived minor-actinides (Np, Am, Cm) transmutation is one of major part of problem of nuclear power ecological safety. The problem of Pu surpluses burning-out adjoins to this problem. Existing and perspective reactor systems could be used for it, but task of optimum organization of the external closed cycle for actinide burner reactor becomes the important aspect of transmutation problem. Since 1992, SSC RIAR has proposed the demonstration program-concept DOVITA (Dry reprocessing, Oxide fuel, Vibropac, Integral, Transmutation of Actinides), which should demonstrate opportunities of new technologies for realization of the optimized fuel cycle for actinide burner reactor. The brief review of study on DOVITA program for 5 years is given in this paper. (J.P.N.)

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

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    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.

  9. Field testing the prototype BNL fan-atomized oil burner

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, R.; Celebi, Y. [Brookhaven National Lab., Upton, NY (United States)

    1995-04-01

    BNL has developed a new oil burner design referred to as the Fan Atomized burner System. The primary objective of the field study was to evaluate and demonstrate the reliable operation of the Fan Atomized Burner. The secondary objective was to establish and validate the ability of a low firing rate burner (0.3-0.4 gph) to fully satisfy the heating and domestic hot water load demands of an average household in a climate zone with over 5,000 heating-degree-days. The field activity was also used to evaluate the practicality of side-wall venting with the Fan Atomized Burner with a low stack temperature (300F) and illustrate the potential for very high efficiency with an integrated heating system approach based on the Fan Atomized Burner.

  10. Fundamental experiments for FBX burner linear fusion reactor core with FBX plasma flow

    International Nuclear Information System (INIS)

    FBX is a production and confinement scheme of a spherical torus that carries a strong plasma current with both toroidal and poloidal components. On the other hand HI-I is a fundamental experiment on moving plasma. With two of them, a new type of fusion reactor scheme FBX-III BURNER (III) is established. In this paper, the fundamental results of the first two types of experiments are introduced to find out a total plasma behavior of the long term project. 9 refs., 5 figs

  11. INITIAL TEST RESULTS OF THE LIMESTONE INJECTION MULTISTAGE BURNER (LIMB) DEMONSTRATION PROJECT

    Science.gov (United States)

    The paper discusses SO2 removal efficiency and low-NOx burner performance obtained during short term tests, as well as the impact of LIMB ash on electrostatic precipitator (ESP) performance at Ohio Edison's Edgewater Station. Project goals are to demonstrate 50% or more SO2 remov...

  12. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  13. 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)

    2012-07-01

    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

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

    International Nuclear Information System (INIS)

    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-ZrO2) 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. Preliminary results

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

    International Nuclear Information System (INIS)

    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 ZrO2 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 ZRO2 based inert matrix fuels, Journal of Nuclear Materials. 361 (2007) 41-51. (authors)

  16. Dilapidation of the TBC system during the Burner Rig Test

    Directory of Open Access Journals (Sweden)

    S Sreenivas

    2015-06-01

    Full Text Available Substrate of Inconel 718 was deposited with a bond coat of nickel cobalt chromium aluminium yttriym (NiCoCrAlY. A top coat of thermal barrier coating of 8% Yttria stabilised zirconia (YSZ was sprayed over the bond coat by an air plasma spray (APS technique by employing standard process parameters. Static oxidation test conducted at 1000 0C and for 120 hours (h revealed that main degradation modes of the TBC system were connected with formation of porous NiAl2O4 oxides in the thermally grown oxide area followed by formation of micro-cracks, delamination of ceramic layer and spallation of ceramic topcoat.

  17. Mathematical modeling and experimental tests of the air jets mixing process in a new prototype of lignite burner for lower NOx emissions

    International Nuclear Information System (INIS)

    In order to decrease the NOx emissions generated by a lignite steam generator, a new prototype burner has been designed. The burner should operate with staggered combustion, to achieve lower temperatures in the first zone of the flame. The paper presents in parallel the results of the mathematical model and experimental tests of the mixing process of primary, secondary and tertiary air-flow jets at environmental temperature. Keywords: mathematical modeling, experimental tests, lignite burner, NOx reduction

  18. The different facilities of the reactor PHENIX for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (author)

  19. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  20. 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)

    2015-11-04

    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

  1. Status of the EC-FP7 Project ARCAS: Comparing the economics of accelerator-driven systems and fast reactors as minor actinide burners

    International Nuclear Information System (INIS)

    The ARCAS Project aims to compare, on a technological and economical basis, accelerator-driven systems and fast reactors as minor actinide burners. It is split into five work packages: the reference scenario definition, the fast reactor system definition, the accelerator-driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarises the status of the project and its five work packages. (authors)

  2. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  3. Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

    International Nuclear Information System (INIS)

    This paper shows the impact of recycling light water reactor (LWR) mixed oxide (MOX) fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radioactivities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burnup increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross-section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10 amounts to 255 years when average burnups are limited to 150 GW.d t-1 (tonne). (orig.)

  4. Development and Test Performance of a Three-Burner Wood-Fired Stove

    International Nuclear Information System (INIS)

    A three-burner wood stove has been designed, constructed and its performance characteristics tested by computing the percentages of wood consumed and the quantity of heat utilized when three pots containing water were heated simultaneously in the three cooking compartments respectively. The result shows that the percentages of wood consumed and the heat energy used were 72.5% and 33.6% respectively. These indicate an improvement on the previous work, which gave the percentages of wood consumed and heat energy utilized to be respectively 61.8% and 10.4% for three-stone wood stove and 79.8% and 19.8% for a two-hole stove. Also the central heating compartment received more heat energy (465.4 KJ) than the side ones (336.3 KJ and 298.2 KJ) after 36 minutes so that the user could discriminate on their application with respect to the heat requirements of what is being cooked

  5. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF3, etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as to

  6. Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

    International Nuclear Information System (INIS)

    The paper shows the impact of recycling LWR-MOX fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radio activities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burn up increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10, amounts to 255 years when average burn ups are limited to 150 GWd t-1. (orig.)

  7. Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    International Nuclear Information System (INIS)

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: 1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, 2) a 'singletier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and 3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall bus-bar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  8. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  9. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Davies, T.E.; Quinn, D.E.; Watson, J.E.

    1986-08-05

    A regenerative burner is described operable in fire and flue modes comprising: a burner shell having first and second internal chambers, the first chamber being disposed on the flame axis of the burner and the second chamber surrounding the radial perimeter of the first chamber; a gas permeable annular regenerative bed separating the first and second chambers such that gas flow between the first and second chambers must travel through the regenerative bed in a generally radial direction with respect to the flame axis; means for supplying combustion air to the second chamber when the burner is in the fire mode and for exhausting the products of combustion from the second chamber when the burner is in the flue mode; and means for supplying fuel in the vicinity of the flame axis for mixing with combustion air to support combustion when the burner is in the fire mode.

  10. Regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Gitman, G.M.

    1990-05-08

    This patent describes a method of combusting fuel in a furnace having a pair of regenerative burners, each burner having a combustion chamber. It comprises: supplying fuel and oxygen alternatively to each burner to create alternating firing burners wherein the oxygen is supplied from two sources providing first and second oxidizing gases having different oxygen concentrations and simultaneously alternating the application of negative pressure to the remaining non-firing burner to recover heat from flue gases exhausted by the regenerative bed of the non-firing burner to be used further to preheat at least part of the oxygen being supplied to the firing burner; mixing the fuel with a fraction of the oxygen under substoichiometric combustion condition to create products of incomplete combustion to form a hot, luminous flame core containing partially pyrolized fuel; and mixing the partially pyrolyzed fuel with a remaining fraction of the oxygen to complete combustion of the pyrolized fuel; and controlling the total flow of fuel and oxygen supplied to each burner to provide each burner with a desired flame stoichiometry.

  11. Oxy-Combustion Burner and Integrated Pollutant Removal Research and Development Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Mark Schoenfield; Manny Menendez; Thomas Ochs; Rigel Woodside; Danylo Oryshchyn

    2012-09-30

    A high flame temperature oxy-combustion test facility consisting of a 5 MWe equivalent test boiler facility and 20 KWe equivalent IPR® was constructed at the Hammond, Indiana manufacturing site. The test facility was operated natural gas and coal fuels and parametric studies were performed to determine the optimal performance conditions and generated the necessary technical data required to demonstrate the technologies are viable for technical and economic scale-up. Flame temperatures between 4930-6120F were achieved with high flame temperature oxy-natural gas combustion depending on whether additional recirculated flue gases are added to balance the heat transfer. For high flame temperature oxy-coal combustion, flame temperatures in excess of 4500F were achieved and demonstrated to be consistent with computational fluid dynamic modeling of the burner system. The project demonstrated feasibility and effectiveness of the Jupiter Oxygen high flame temperature oxy-combustion process with Integrated Pollutant Removal process for CCS and CCUS. With these technologies total parasitic power requirements for both oxygen production and carbon capture currently are in the range of 20% of the gross power output. The Jupiter Oxygen high flame temperature oxy-combustion process has been demonstrated at a Technology Readiness Level of 6 and is ready for commencement of a demonstration project.

  12. Lead coolant for the fast reactor-burner with a hard neutron spectrum

    International Nuclear Information System (INIS)

    The possibility of the efficiency increasing of the junior actinides burn-up on the account of their fission in the fast reactor by using the Pb-208 lead isotope as a coolant is considered. The calculation relative values of the fission reduction: rates and capture the neutrons by the Np-237, Am-241 and Am-243 nuclei in the fast reactor blanket for different coolants: sodium, lead, lead of natural composition and lead with 100%-Pb-208 content are presented. The radioactivity, induced during one year of the reactor operation and its decay during the time up to 1000 years is calculated for the lead coolants of various isotopic composition

  13. Concept of fast reactors-plutonium burners and their fuel cycle

    International Nuclear Information System (INIS)

    In this report the concept is considered of a closed fuel cycle of nuclear power, consisting of thermal and fast reactors, providing utilization of practically all actinides produced in the nuclear power. The major calculation results and the ways to form fast reactors cores for effective actinides burning are presented. The existing limitations on the fresh fuel composition by heat release and radiation characteristics are given. The calculation studies results on the fuel cycle characteristics at repeated fuel recycling in a system of VVER-1000 and BN-800 types reactors are presented. The calculations were carried out for different type cores of the BN-800 reactor -with oxide fuel of increased enrichment and with fuel without uranium-238, varying the burn-out level and the decay time of spent fuel. (author)

  14. 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

    2001-02-01

    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

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

    International Nuclear Information System (INIS)

    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

  16. The IBR-2 test reactor

    International Nuclear Information System (INIS)

    Major design criteria, specifications and potential fields of application of the IBR-2 pulsed test reactor (now under construction in Dubna, USSR) are described. The pulsed power bursts will be due to fast periodic reactivity changes by a rotating reflector. The frequency of approximately 100 μs pulsed may be 5, 12.5 or 50 Hz. The IBR-2 reactor will be mostly profitable for slow neutron experiments when investigating solids, nuclei or neutrons themselves using spectroscopic methods. Due to the high peak flux of thermal neutrons (1016-1017 n/cm2xs) the reactor will be superior (for the sort of experiments) to the currently operating SM-2 and HFR high flux steady-state test reactors for many times

  17. Cooperative Russian-French experiment on plutonium-enriched fuels for fast burner reactor

    International Nuclear Information System (INIS)

    Various kinds of nuclear fuels with an increased plutonium content are under study according to the program including three stages: fabrication, irradiation in BOR-60 reactor, post-irradiation examination. Flowsheets for fabricating pelletized and vibrocompacted fuels of UPu0.45O2, UPu0.45N, UPu0.6N, PuN + ZrN, PuO2 + MgO are presented along with basic fuel properties. The irradiation of oxide fuel is carried out in an individual irradiation device at rated maximum temperature of the fuel at the beginning of irradiation equal to 2100 deg C. The irradiation of nitride fuel and the fuel based on inert matrices is performed in the other device with the aim of limitation of maximum temperature by the value of 1550 deg C. The duration of irradiation for all fuel types constitutes 750 EFPD. Fuel element charge in Bor-60 reactor core was realized in 2000

  18. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Kent Alan [ORNL; Shropshire, David E. [Idaho National Laboratory (INL)

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a

  19. Neutronic design of a plutonium-thorium burner small nuclear reactor

    International Nuclear Information System (INIS)

    A small nuclear reactor using thorium and plutonium fuel has been designed from the neutronic point of view. The thermal power of the reactor is 150 MWth and it is proposed to be used to supply electricity in an island in Indonesia. Thorium and plutonium fuel was chosen because in recent years the thorium fuel cycle is one of the promising ways to deal with the increasing number of plutonium stockpiles, either from the utilization of uranium fuel cycle or from nuclear weapon dismantling. A mixed fuel of thorium and plutonium will not generate the second generation of plutonium which will be a better way to incinerate the excess plutonium compared with the MOX fuel. Three kinds of plutonium grades which are the reactor grade (RG), weapon grade (WG), and spent fuel grade (SFG) plutonium, were evaluated as the thorium fuel mixture in the 17x17 Westinghouse PWR Fuel assembly. The evaluated parameters were the multiplication factor, plutonium depletion, fissile buildup, neutron spectrum, and temperature reactivity feedback. An optimization was also done to increase the plutonium depletion by changing the Moderator to Fuel Ratio (MFR). The computer codes TRITON (coupled NEWT and ORIGEN-S) in SCALE version 6 were used as the calculation tool for this assembly level. From the evaluation and optimization of the fuel assembly, the whole core was designed. The core was consisted of 2 types of thorium fuel with different plutonium grade and it followed the checkerboard loading pattern. A new concept of enriched burnable poison was also introduced to the core. The core life is 6.4 EFPY or 75 GWd/MTHM. It can burn up to 58% of its total mass of initial plutonium. VENTURE was used as the calculation tool for the core level

  20. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  1. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  2. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  3. Testing and Modeling Fuel Regression Rate in a Miniature Hybrid Burner

    OpenAIRE

    Luciano Fanton; Christian Paravan; Luigi T. De Luca

    2012-01-01

    Ballistic characterization of an extended group of innovative HTPB-based solid fuel formulations for hybrid rocket propulsion was performed in a lab-scale burner. An optical time-resolved technique was used to assess the quasisteady regression history of single perforation, cylindrical samples. The effects of metalized additives and radiant heat transfer on the regression rate of such formulations were assessed. Under the investigated operating conditions and based on phenomenological models ...

  4. Testing 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

    2007-07-01

    Combined heat and power (CHP) or cogeneration involves the generation of electricity in addition to the productive use of waste heat from the combustion process using the same primary fuel. An alternative to combined electrical power and heat generation is a micro-cogeneration unit which uses a micro-turbine as a prime mover. This type of unit is expected to result in a shift from large and centralized plants to smaller, more economical on-site generation plants. This paper presented a new low nitrogen oxide (NOx) wire-mesh duct burner (WMDB) for the development of a more efficient micro-cogeneration unit. In order to increase its thermal output, the low NOx WMDB was designed, built and integrated for evaluation with the Ingersol-Rand 70 kw micro-cogeneration unit. The wire-mesh burner had a conical shape and was manufactured by ACOTECH. The paper also discussed the advantages of micro-CHP units which are more attractive to building owners, retail establishments, commercial and light industrial facilities. Advantages include quality of the power supply; more economical, cleaner power; and the addition of new capacity without new transmission lines. It was concluded that low levels of emission were achieved with the development of a low NOx wire-mesh duct burner for a micro-cogeneration plant. 2 refs., 5 figs.

  5. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  6. Development and testing of the pore burner technology for oil burners. Final report; Entwicklung und Erprobung der Porenbrennertechnik fuer Oelbrenner. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Durst, F.; Trimis, D.; Wawrzinek, K.; Koehne, H.; Lucka, K.; Rudolphi, I.; Hatzfeld, O.; Volkert, J.; Rutsche, A.; Adler, J.; Standke, G.; Haase, F.; Krueger, K.; Kuechen, C.

    2001-11-01

    The application of the pore burner technology in oil burners was investigated. Together with a new concept for oil-fuelled high efficiency boilers, this technology opens up a vast potential for energy conservation and pollutant reduction. [German] Der Waermebedarf von Wohneinheiten nimmt, flankiert durch Vorgaben des Gesetzgebers, in Zukunft weiter ab. Parallel dazu werden die Grenzwerte fuer die maximal zulaessigen Schadgasemissionen der Heizanlagen verschaerft und die emissionsintensiven und im intermittierenden Betrieb bei Teillast sehr haeufigen Start/Stop-Betriebsphasen konventioneller Oel-Heizsysteme strenger bewertet. Ziel dieses Vorhabens ist es, die fuer die Verbrennung gasfoermiger Brennstoffe bereits erfolgreich demonstrierten Vorteile der Porenbrennertechnik (sehr niedrige Schadstoffemissionen, aeusserst breiter Bereich der Leistungsmodulation bis 1:20, hohe Energiedichte und damit kleine Baugroesse, minimale Geraeuschemission) auch fuer die Verbrennung von Heizoel nutzbar zu machen. In Verbindung mit einem neuen Konzept fuer die Oel-Brennwerttechnik erschliesst diese Technologie ein hohes Einsparpotential hinsichtlich Energieverbrauch und Schadstoffemissionen. (orig.)

  7. Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

    International Nuclear Information System (INIS)

    other self-actuating reactivity devices have been investigated in the past, but more as safety shutdown features than for continuous feedback, for example, 1. the gas expansion module tested in the Fast Flux Test Facility (FFTF), in which core pressure drop compresses a gas bubble into the upper region of a vertical in-core through-tube. In a loss-of-flow (LOF) incident, the bubble expands to void the tube and remove reactivity by increasing leakage. 2. a fluidized bed of tantalum neutron poison balls, held out of core by coolant flow, which drops into a channel through the core during an LOF sequence; 3. a magnetic latch that exceeds its curie point in the temperature rise following a transient overpower event, and thereby allows a control absorber to fall into the core. The RFD is not a substitute for such other devices, nor does it avoid the need for conventional electro-mechanically driven control rods for regulation and shutdown. In fact, the latter will have to be enhanced to compensate for the automatic RFD withdrawal following scram. In summary, an RFD has been designed to compensate for some of the reactivity feedback shortcomings of actinide burner reactors. In principle, similar devices could also be used in other reactor types. (authors)

  8. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  9. FASTER test reactor preconceptual design report summary

    International Nuclear Information System (INIS)

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  10. FIELD EVALUATION OF LOW-EMISSION COAL BURNER TECHNOLOGY ON UTILITY BOILERS VOLUME II. SECOND GENERATION LOW-NOX BURNERS

    Science.gov (United States)

    The report describes tests to evaluate the performance characteristics of three Second Generation Low-NOx burner designs: the Dual Register burner (DRB), the Babcock-Hitachi NOx Reducing (HNR) burner, and the XCL burner. The three represent a progression in development based on t...

  11. 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)

    None

    1977-10-01

    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.

  12. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  13. Unirradiated high temperature reactor fuel element head-end reprocessing tests

    International Nuclear Information System (INIS)

    For several years, the United States and the Federal Republic of Germany (FRG) have engaged in a successful cooperative program to develop high temperature gas-cooled reactor (HTGR) fuel cycle technology. Recent tests in reprocessing pilot plant facilities at General Atomic Company have demonstrated the feasibility of performing HTGR head-end unit operations for both spherical (German) and block-type (American) fuel elements in a single process line. Because of an unexpected high fines generation and elutriation rate, extended fluidized bed primary burning of FRG fuel material was impossible to accomplish with the burner system and operating procedures optimized for U.S. fuel burning. Operational modification, including startup with a carbon-poor bed and reduction of the fluidizing velocity, resulted in dramatic improvements in FRG fuel-burning behavior and allowed extended processing campaigns. Additional modifications to the fines recycle system and burner are recommended to optimize the system for processing of FRG fuels

  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)

    1997-09-01

    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. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  16. Optimisation of efficiency and emissions in pellet burners

    International Nuclear Information System (INIS)

    There is a trade-off between the emissions of nitrogen oxides (NOx) and of unburnt hydrocarbons and carbon monoxide (OGC and CO). Decreasing the excess air results in lower NOx emission but also increased emission of unburnt. The efficiency increases, as the excess air is decreased until the losses due to incomplete combustion become too high. The often-high NOx emission in today's pellet burners can be significantly reduced using well-known techniques such as air staging. The development of different chemical sensors is very intensive and recently sensors for CO and OGC have been introduced on the market. These sensors may, together with a Lambda sensor, provide efficient control for optimal performance with respect to emissions and efficiency. In this paper, results from an experimental parameter study in a modified commercial burner, followed by Chemkin simulations with relevant input data and experiments in a laboratory reactor and in a prototype burner, are summarised. Critical parameters for minimisation of NOx emission from pellet burners are investigated in some detail. Also, results from tests of a new sensor for unburnt are reported. In conclusion, relatively simple design modifications can significantly decrease NOx emission from today's pellet burners

  17. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  18. Testing and Modeling Fuel Regression Rate in a Miniature Hybrid Burner

    Directory of Open Access Journals (Sweden)

    Luciano Fanton

    2012-01-01

    Full Text Available Ballistic characterization of an extended group of innovative HTPB-based solid fuel formulations for hybrid rocket propulsion was performed in a lab-scale burner. An optical time-resolved technique was used to assess the quasisteady regression history of single perforation, cylindrical samples. The effects of metalized additives and radiant heat transfer on the regression rate of such formulations were assessed. Under the investigated operating conditions and based on phenomenological models from the literature, analyses of the collected experimental data show an appreciable influence of the radiant heat flux from burnt gases and soot for both unloaded and loaded fuel formulations. Pure HTPB regression rate data are satisfactorily reproduced, while the impressive initial regression rates of metalized formulations require further assessment.

  19. RF torch discharge combined with conventional burner

    International Nuclear Information System (INIS)

    The design of the combined flame-rf-plasma reactor and experimental examination of this reactor are presented. For the determination of the temperature in different parts of the combined burner plasma the methods of emission spectroscopy were used. The temperatures measured in the conventional burner reach the maximum temperature 1900 K but in the burner with the superimposed rf discharge the neutral gas temperature substantially increased up to 2600 K but also the plasma volume increases substantially. Consequently, the resident time of reactants in the reaction zone increases

  20. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  1. Process development report: 0.40-m primary burner system

    International Nuclear Information System (INIS)

    Fluidized bed combustion is required in reprocessing the graphite-based fuel elements from high-temperature gas-cooled reactor (HTGR) cores. This burning process requires combustion of beds containing both large particles and very dense particles, and also of fine graphite particles which elutriate from the bed. This report documents the successful long-term operation of the 0.40-m primary burner in burning crushed fuel elements. The 0.40-m system operation is followed from its first short heatup test in September 1976 to a > 40-h burning campaign that processed 20 LHTGR blocks in September 1977. The 0.40-m perforated conical gas distributor, scaled up from the 0.20-m primary burner, has proven reliable in safely burning even the largest, densest adhered graphite/fuel particle clusters originating from the crushing of loaded fuel elements. Such clusters had never been fed to the 0.20-m system. Efficient combustion of graphite fines using the pressurized recycle technique was demonstrated throughout the long-duration operation required to reduce a high carbon fresh feed bed to a low carbon particle bed. Again, such operation had never been completed on the 0.20-m system from which the 0.40-m burner was scaled. The successful completion of the tests was due, in part, to implementation of significant equipment revisions which were suggested by both the initial 0.40-m system tests and by results of ongoing development work on the 0.2-m primary burner. These revisions included additional penetrations in the burner tube side-wall for above-bed fines recycle, replacement and deletion of several metal bellows with bellows of more reliable design, and improvements in designs for burner alignment and feeder mechanisms. 76 figures, 8 tables

  2. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  3. Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

    International Nuclear Information System (INIS)

    Research highlights: → Double-strata fuel cycle has a potential to minimize transuranics mass in Europe. → European Minor Actinides legacy can be reduced down to 0 before the end of century. → 40% higher capacity needed to burn MA for fast critical reactor then for EFIT fleet. → Na cooled fast reactor cores with high content of MA and low CR have been assessed. → Fast critical and ADS-EFIT reactors show comparable MA transmutation performance. - Abstract: In the frame of Partitioning and Transmutation (P and T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a

  4. Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead-bismuth

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed

  5. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  6. Fundamental studies on porous flame reactors for minimizing pollutant emissions of premix burners. Continued report; Grundlagenuntersuchungen an poroesen Flammenreaktoren zur Minimierung von Schadgasemissionen bei der vorgemischten Verbrennung. Fortsetzungsbericht

    Energy Technology Data Exchange (ETDEWEB)

    Durst, F.; Moessbauer, S.

    2001-01-31

    The report summarizes investigations of effective heat transport processes inside highly porous solid structures. These heat transport processes are of decisive importance for the pore burner technology developed at Erlangen-Nuremberg University. A test stand was set up for recording the two-dimensional temperature field of cross-flowed solid structures. [German] Der vorliegende Bericht fasst Arbeiten zusammen, die sich mit der Bestimmung effektiver Waermetransportvorgaenge im Inneren von hochporoesen Festkoerperstrukturen befassen. Diese Waermetransportvorgaenge sind entscheidend fuer die Vorteile der am Lehrstuhl fuer Stroemungsmechanik der Friedrich-Alexander-Universitaet Erlangen-Nuernberg entwickelten Porenbrennertechnologie. Um diese Vorteile besser zu nutzen und um diese neuartige Technologie weiter verbessern zu koennen, ist es erforderlich, dass die ablaufenden Waermetransportvorgaenge im Inneren von hochporoesen Strukturen im Detail verstanden werden. Zu diesem Zweck wurde ein Versuchsstand erstellt, mit dem das zweidimensionale Temperaturfeld von durchstroemten Festkoerperstrukturen erfasst werden kann. (orig.)

  7. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  8. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  9. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  10. ESTABLISHMENT OF DESIGN CRITERIA FOR OPTIMUM BURNERS FOR APPLICATION TO HEAVY FUEL FIRED PACKAGE BOILERS. VOLUME 2. PILOT SCALE TESTS

    Science.gov (United States)

    The report gives results of a research program to develop low-NOx heavy oil burners for application to industrial package boilers. Volume I documents Phase 1 of the program, bench scale studies which defined optimum conditions for two-stage combustion. The information led to a co...

  11. ESTABLISHMENT OF DESIGN CRITERIA FOR OPTIMUM BURNERS FOR APPLICATION TO HEAVY FUEL FIRED PACKAGE BOILERS. VOLUME 1. LABORATORY SCALE TESTS

    Science.gov (United States)

    The report gives results of a research program to develop low-NOx heavy oil burners for application to industrial package boilers. Volume I documents Phase 1 of the program, bench scale studies which defined optimum conditions for two-stage combustion. The information led to a co...

  12. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  13. Short-term corrosion testing in a burner rig with oxy-fuel and conventional firing

    International Nuclear Information System (INIS)

    As part of the European ENCAP project, a test rig has been modified by IVD (Institute of Process - Engineering and Power Plant Technology), Stuttgart, Germany to compare oxy-fuel firing with conventional firing. A broad spectrum of alloys from low alloyed steels to nickel alloys were exposed on a cooled probe in this rig at a metal temperature of 630 C for up to 40 hours with lignite from the Lausitz region as the fuel, and for 26 hours with bituminous coal (Kleinkopje) as fuel. For lignite firing, both exposure with oxy-fuel firing with recirculation of flue gas and conventional firing has been conducted to compare the corrosion attack and deposit composition. Only oxy-fuel tests were conducted with the bituminous coal. For the lignite fuel, the deposit composition from oxy-fuel and conventional firing was similar, and consisted of calcium sulphate (anhydrite) and iron oxide (hematite). The corrosion attack for the different alloys was also similar for both types of combustion. The corrosion attack was oxidation with some sulfidation/sulphation at the oxide-metal interface. In general, the thickness of the corrosion product decreased for the higher alloyed steels. The ferritic steels formed a two layered oxide. The high nickel containing alloy Hastelloy C-2000 showed a surprisingly high corrosion rate with internal attack and an outer nickel oxide. The Kanthal APM and Nimonic 263 alloys had the best oxide with even surface coverage. For the bituminous coal test, there was very little deposit on the specimens, and the deposit was rich in aluminium, silicon, oxygen and iron with lesser amounts of calcium and sulphur. The corrosion attack for the different alloys was similar to that observed for lignite where Kanthal APM had the most protective oxide coverage. Generally the mass gain rates are similar between the different tests for the same material. The main findings from these short term tests are that alumina forming alloys and super austenitic stainless steels

  14. Short-term corrosion testing in a burner rig with oxy-fuel and conventional firing

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, M. [Vattenfall Heat Nordic/DTU Mekanik (Denmark); Hjornhede, A. [Vattenfall Power Consultant AB (Sweden); Gerhardt, A. [Institute of Process-Engineering and Power Plant Technology (IVD)/ Vattenfall Research and Development (Germany)

    2009-07-01

    As part of the European ENCAP project, a test rig has been modified by IVD (Institute of Process - Engineering and Power Plant Technology), Stuttgart, Germany to compare oxy-fuel firing with conventional firing. A broad spectrum of alloys from low alloyed steels to nickel alloys were exposed on a cooled probe in this rig at a metal temperature of 630 C for up to 40 hours with lignite from the Lausitz region as the fuel, and for 26 hours with bituminous coal (Kleinkopje) as fuel. For lignite firing, both exposure with oxy-fuel firing with recirculation of flue gas and conventional firing has been conducted to compare the corrosion attack and deposit composition. Only oxy-fuel tests were conducted with the bituminous coal. For the lignite fuel, the deposit composition from oxy-fuel and conventional firing was similar, and consisted of calcium sulphate (anhydrite) and iron oxide (hematite). The corrosion attack for the different alloys was also similar for both types of combustion. The corrosion attack was oxidation with some sulfidation/sulphation at the oxide-metal interface. In general, the thickness of the corrosion product decreased for the higher alloyed steels. The ferritic steels formed a two layered oxide. The high nickel containing alloy Hastelloy C-2000 showed a surprisingly high corrosion rate with internal attack and an outer nickel oxide. The Kanthal APM and Nimonic 263 alloys had the best oxide with even surface coverage. For the bituminous coal test, there was very little deposit on the specimens, and the deposit was rich in aluminium, silicon, oxygen and iron with lesser amounts of calcium and sulphur. The corrosion attack for the different alloys was similar to that observed for lignite where Kanthal APM had the most protective oxide coverage. Generally the mass gain rates are similar between the different tests for the same material. The main findings from these short term tests are that alumina forming alloys and super austenitic stainless steels

  15. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  16. Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

    Energy Technology Data Exchange (ETDEWEB)

    Salvatores, Massimo [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States)], E-mail: massimo.salvatores@cea.fr; Chabert, Christine [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Fazio, Concetta [Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe (Germany); Hill, Robert [Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States); Peneliau, Yannick; Slessarev, Igor [CEA Cadarache, 13108 St-Paul-Lez-Durance (France); Yang, Won Sik [Argonne National Laboratory, NE Division, Argonne, IL 60439 (United States)

    2009-10-15

    Partitioning and Transmutation (P and T) strategies assessment and implementation play a key role in the definition of advanced fuel cycles, in order to insure both sustainability and waste minimization. Several options are under study worldwide, and their impact on core design and associated fuel cycles are under investigation, to offer a rationale to down selection and to streamline efforts and resources. Interconnected issues like fuel type, minor actinide content, conversion ratio values, etc. need to be understood and their impact quantified. Then, from a practical point of view, studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for more detailed studies. In this paper, it is presented an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the specific issue of the features of the fuel cycle parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different ratios of Pu over minor actinide (Pu/MA) in the fuel feed. As an example of the potential impact of such analysis, it is shown that for cores with CR below {approx}0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles.

  17. Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

    International Nuclear Information System (INIS)

    Partitioning and Transmutation (P and T) strategies assessment and implementation play a key role in the definition of advanced fuel cycles, in order to insure both sustainability and waste minimization. Several options are under study worldwide, and their impact on core design and associated fuel cycles are under investigation, to offer a rationale to down selection and to streamline efforts and resources. Interconnected issues like fuel type, minor actinide content, conversion ratio values, etc. need to be understood and their impact quantified. Then, from a practical point of view, studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for more detailed studies. In this paper, it is presented an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the specific issue of the features of the fuel cycle parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different ratios of Pu over minor actinide (Pu/MA) in the fuel feed. As an example of the potential impact of such analysis, it is shown that for cores with CR below ∼0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles.

  18. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  19. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  20. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  1. Thermionic cogeneration burner assessment

    International Nuclear Information System (INIS)

    Both electric power and high-temperature flue gas can be cogenerated by combining a furnace burner with thermionic converters, forming a thermionic cogeneration burner. To assess the performance and cost of such a burner, a one-for-one replacement, bolt-on burner, which could be used in most industrial applications, was designed in detail. It was analyzed and parametric performance data was derived from a mathematical model. Details of the design analysis, as well as an economic evaluation of installed cost ($/kW) and internal rate-of-return, are presented

  2. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  3. In-reactor testing of ionic thermometers

    International Nuclear Information System (INIS)

    Ionic thermometers have been tested in a nuclear reactor with attention to the steepness of the ionic conductivity jump and the influence of a glass container on the accuracy of the temperature measurements. It was found that, at the neutron fluxes up to 1.5 x 1018 m-2 s-1 (thermal) and 3 x 1018 m-2 s-1 (fast) in a light water reactor, the change of conductivity jump slope is negligible or nil for an ionic thermometer filled by HgI2, i.e., at 256.0 +- 0.2 0C. The need to use boron-free glass was confirmed. The impact on the accuracy of the temperature point indication in a nuclear reactor core is discussed, as well as obvious inertness of the melting process mechanism to the intense irradiation field

  4. MINIMIZATION OF NO EMISSIONS FROM MULTI-BURNER COAL-FIRED BOILERS; FINAL

    International Nuclear Information System (INIS)

    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

  5. MINIMIZATION OF NO EMISSIONS FROM MULTI-BURNER COAL-FIRED BOILERS

    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

    2002-01-01

    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

  6. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

  7. Process development report: 0. 20-m secondary burner system

    Energy Technology Data Exchange (ETDEWEB)

    Rickman, W.S.

    1977-09-01

    HTGR fuel reprocessing consists of crushing the spent fuel elements to a size suitable for burning in a fluidized bed to remove excess graphite; separating, crushing, and reburning the fuel particles to remove the remainder of the burnable carbon; dissolution and separation of the particles from insoluble materials; and solvent extraction separation of the dissolved uranium and thorium. Burning the crushed fuel particles is accomplished in a secondary burner. This is a batch fluidized-bed reactor with in-vessel, off-gas filtration. Process heat is provided by an induction heater. This report documents operational tests performed on a commercial size 0.20-m secondary burner using crushed Fort St. Vrain type TRISO fuel particles. Analysis of a parametric study of burner process variables led to recommending lower bed superficial velocity (0.8 m/s), lower ignition temperature (600/sup 0/C), lower fluid bed operating temperature (850/sup 0/C), lower filter blowback frequency (1 cycle/minute), and a lower fluid bed superficial velocity during final bed burnout (0.45 m/s).

  8. Process development report: 0.20-m secondary burner system

    International Nuclear Information System (INIS)

    HTGR fuel reprocessing consists of crushing the spent fuel elements to a size suitable for burning in a fluidized bed to remove excess graphite; separating, crushing, and reburning the fuel particles to remove the remainder of the burnable carbon; dissolution and separation of the particles from insoluble materials; and solvent extraction separation of the dissolved uranium and thorium. Burning the crushed fuel particles is accomplished in a secondary burner. This is a batch fluidized-bed reactor with in-vessel, off-gas filtration. Process heat is provided by an induction heater. This report documents operational tests performed on a commercial size 0.20-m secondary burner using crushed Fort St. Vrain type TRISO fuel particles. Analysis of a parametric study of burner process variables led to recommending lower bed superficial velocity (0.8 m/s), lower ignition temperature (6000C), lower fluid bed operating temperature (8500C), lower filter blowback frequency (1 cycle/minute), and a lower fluid bed superficial velocity during final bed burnout

  9. Process development report: 0. 40-m primary burner system. [Spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Young, D.T.

    1978-04-01

    Fluidized bed combustion is required in reprocessing the graphite-based fuel elements from high-temperature gas-cooled reactor (HTGR) cores. This burning process requires combustion of beds containing both large particles and very dense particles, and also of fine graphite particles which elutriate from the bed. This report documents the successful long-term operation of the 0.40-m primary burner in burning crushed fuel elements. The 0.40-m system operation is followed from its first short heatup test in September 1976 to a > 40-h burning campaign that processed 20 LHTGR blocks in September 1977. The 0.40-m perforated conical gas distributor, scaled up from the 0.20-m primary burner, has proven reliable in safely burning even the largest, densest adhered graphite/fuel particle clusters originating from the crushing of loaded fuel elements. Such clusters had never been fed to the 0.20-m system. Efficient combustion of graphite fines using the pressurized recycle technique was demonstrated throughout the long-duration operation required to reduce a high carbon fresh feed bed to a low carbon particle bed. Again, such operation had never been completed on the 0.20-m system from which the 0.40-m burner was scaled. The successful completion of the tests was due, in part, to implementation of significant equipment revisions which were suggested by both the initial 0.40-m system tests and by results of ongoing development work on the 0.2-m primary burner. These revisions included additional penetrations in the burner tube side-wall for above-bed fines recycle, replacement and deletion of several metal bellows with bellows of more reliable design, and improvements in designs for burner alignment and feeder mechanisms. 76 figures, 8 tables.

  10. Design and construction of thermionic cogeneration burner module

    International Nuclear Information System (INIS)

    The thermionic cogeneration burner module is a high temperature burner equipped with thermionic converters. A demonstration of a thermionic cogeneration system is under way. In this demonstration a hot oil heater (used in various industrial processes) was equipped with a thermionic cogeneration burner module. This module contained converters that were connected in series to produce approximately 180 watts at 2.4 volts. The system is now undergoing preliminary testing. It is expected that additional test results will be available in the fall

  11. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  12. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  13. Combustor burner vanelets

    Energy Technology Data Exchange (ETDEWEB)

    Lacy, Benjamin (Greer, SC); Varatharajan, Balachandar (Loveland, OH); Kraemer, Gilbert Otto (Greer, SC); Yilmaz, Ertan (Albany, NY); Zuo, Baifang (Simpsonville, SC)

    2012-02-14

    The present application provides a burner for use with a combustor of a gas turbine engine. The burner may include a center hub, a shroud, a pair of fuel vanes extending from the center hub to the shroud, and a vanelet extending from the center hub and/or the shroud and positioned between the pair of fuel vanes.

  14. Catalyzed Ceramic Burner Material

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, Amy S., Dr.

    2012-06-29

    period in accomplishing these objectives. Our work in the area of Pd-based, methane oxidation catalysts has led to the development of highly active catalysts with relatively low loadings of Pd metal using proprietary coating methods. The thermal stability of these Pd-based catalysts were characterized using SEM and BET analyses, further demonstrating that certain catalyst supports offer enhanced stability toward both PdO decomposition and/or thermal sintering/growth of Pd particles. When applied to commercially available fiber mesh substrates (both metallic and ceramic) and tested in an open-air burner, these catalyst-support chemistries showed modest improvements in the NOx emissions and radiant output compared to uncatalyzed substrates. More significant, though, was the performance of the catalyst-support chemistries on novel media substrates. These substrates were developed to overcome the limitations that are present with commercially available substrate designs and increase the gas-catalyst contact time. When catalyzed, these substrates demonstrated a 65-75% reduction in NOx emissions across the firing range when tested in an open air burner. In testing in a residential boiler, this translated into NOx emissions of <15 ppm over the 15-150 kBtu/hr firing range.

  15. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Krystyník, Pavel; Beneš, Ondřej; Klusoň, Petr; Šolcová, Olga

    Praha: Česká společnost průmyslové chemie, 2015, s. 30 /p104./. ISBN 978-80-86238-73-9. [mezinárodní chemicko-technologická konference (ICCT 2015) /3./. Mikulov (CZ), 13.04.2015-15.04.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * heat transfer * testing Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  16. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Beneš, Ondřej; Hanková, Libuše; Klusoň, Petr; Šolcová, Olga

    Bratislava: Slovak Society of Chemical Engineering, 2015 - (Markoš, J.), s. 40 ISBN 978-80-89475-14-8. [International Conference of Slovak Society of Chemical Engineering /42./. Tatranské Matliare (SK), 25.05.2015-29.05.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * testing * partial oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  17. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  18. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  19. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34∼38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc

  20. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  1. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  2. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  3. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    In this work, the results of frequency response tests conducted on ITU TRIGA Reactor are presented. To conduct the experiments, a special 'micro control rod' and its submersible stepping-motor drive mechanism was designed and constructed. The experiments cover a frequency range of 0.002 - 2 Hz., and 0.02, 4, 200 kW nominal power levels. Zero-power and at-power reactivity to % power transfer functions are presented as gain, and phase shift vs. frequency diagrams. Low power response is in close agreement with the point reactor zero-power transfer function. Response at 200 kW is studied with the help of a Nyquist diagram, and found to be stable. An elaboration on the main features of the feedback mechanism is also given. Power to reactivity feedback was measured to be just about 1.5 cent / % power change. (authors)

  4. Gas/particle flow characteristics of two swirl burners

    International Nuclear Information System (INIS)

    A three-component particle-dynamics anemometer is used to measure, in the near-burner region, the characteristics of gas/particle two-phase flows with a centrally fuel rich swirl coal combustion burner and enhanced ignition-dual register burner, on a gas/particle two-phase test facility. Velocities, RMS velocities, particle mean diameters and particle volume flux profiles were obtained. For the centrally fuel rich burner, particles penetrate the central recirculation zone partially, and are then deflected radially. For the enhanced ignition-dual register burner, particles completely penetrate the central recirculation zone. Compared with the enhanced ignition-dual register burner, in the same cross-section, the particle volume flux peak value for the centrally fuel rich burner is much closer to the chamber axis and much larger near the chamber axis. In six cross-sections from x/d = 0.3 to 2.5, the particle volume flux in the central recirculation zone for the centrally fuel rich burner is much larger than that for the enhanced ignition-dual register burner. For the centrally fuel rich burner, most of bigger particles are resident in the region near the chamber axis and the residence time is prolonged. The influence of gas/particle flow characteristics on combustion has been analyzed.

  5. Fuel irradiation test plan at the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Development of high performance fuels, which enables burnup extension and high duty uses of light water reactors (LWRs) by means of power up rates and flexible operating cycles, is one of key technical issues for extending the uses for longer periods. Introduction of new design fuel rods with new cladding alloys and wider utilization of mixed oxide fuels is expected in Japan. Fuel irradiation tests for development and safety demonstration are quite important, in order to realize theses progress. Operational management on water chemistry, minimizing the long term degradation of reactor components, could have unfavorable influence on the integrity of the fuel rods. Japanese government and the Japan Atomic Energy Agency have decided to re new the Japan Materials Testing Reactor (JMTR) and to install new test rigs, in order to play an active role solving the issues on the development and the safety of the fuel and the plant aging. Fuel integrity under abnormal transient conditions will be investigated using a special capsule type test rig, which has its own power control system under simulated LWR cooling conditions. Water loops for simulation of high duty operation, e.g. high power, high burnup and high rod internal pressure conditions, are proposed for the development and safety examination of the high performance fuels. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor and loss of coolant accident tests in hot laboratories would provide a comprehensive data for safety evaluation and design progress of the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients

  6. FIELD EVALUATION OF LOW-EMISSIONS COAL BURNER TECHNOLOGY ON UTILITY BOILERS; VOLUME II. SECOND GENERATION LOW-NOX BOILERS

    Science.gov (United States)

    The report describes tests to evaluate the performance characteristics of three Second Generation Low-NOx burner designs: the Dual Register burner (DRB), the Babcock-Hitachi NOx Reducing (HNR) burner, and the XCL burner. The three represent a progression in development based on t...

  7. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  8. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  9. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart

  10. A heated chamber burner for atomic absorption spectroscopy.

    Science.gov (United States)

    Venghiattis, A A

    1968-07-01

    A new heated chamber burner is described. The burner is of the premixed type, and burner heads of the types conventionally used in atomic absorption may be readily adapted to it. This new sampling system has been tested for Ag, Al, Ca, Cu, Fe, Mg, Mn, Ni, Pb, Si, Ti, and Zn in aqueous solutions. An improvement of the order of ten times has been obtained in sensitivity, and in detection limits as well, for the elements determined. Interferences controllable are somewhat more severe than in conventional burners but are controllable. PMID:20068792

  11. CHP Integrated with Burners for Packaged Boilers

    Energy Technology Data Exchange (ETDEWEB)

    Castaldini, Carlo; Darby, Eric

    2013-09-30

    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

  12. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    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 ampersand 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

  13. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  14. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  15. Relationship between core size, coolant choice, fuel type, and neutron flux in a fast irradiation test reactor

    International Nuclear Information System (INIS)

    Currently, the United States has no domestic capability for large volume irradiation testing in a fast-spectrum system to support the development of advanced fuels or materials. The recently-proposed Global Nuclear Energy Partnership includes provisions for a sodium-cooled Advanced Burner Test Reactor which could provide this testing capability. In addition to sodium, lead-bismuth eutectic and helium coolants are being considered for future energy systems. In this paper, sodium, lead-bismuth eutectic, and helium-cooled systems are evaluated to determine the impact on fast flux and fuel enrichment resulting from varying core diameter, fuel volume fraction, fuel type, and coolant. While fast flux is most strongly influenced by core diameter at fixed power, fuel enrichment is a more complicated function of all four parameters. In the end, the combination of high fast flux and low enrichment can best be achieved by a sodium-cooled system. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  17. Thermal Characteristics of Heating-furnace with Regenerative Burner

    OpenAIRE

    HUA, Jianshe; Li, Xiaoming; Kawabata, Nobuyoshi

    2005-01-01

    Thermal characteristics between the heating-furnace with regenerative burner and the classical triple-fired continuous furnace by heat balance testing for two billet steel heating-furnace at the same billet steel heating have been analyzed. In addition, the operating principle, the thermal characteristics and the effect of energy saving for heating-furnace with regenerative burner are introduced.

  18. Post reactor researches of fuel pins, tested under alternating NEMF reactor functioning modes

    International Nuclear Information System (INIS)

    Changing of rod ceramic fuel pins state under their exploitation conditions changing influence at alternating of three-mode nuclear energy-moving facility reactor functioning has been examined. There are presented the results of researches of fuel pins, tested in the reactor IRGIT and RA, firstly under moving mode, then - under energy mode of minor power of NEMF reactor. (author)

  19. Emission characteristics of a novel low NOx burner fueled by hydrogen-rich mixtures with methane

    OpenAIRE

    Dutka, Marcin Damian; Ditaranto, Mario; Løvås, Terese

    2015-01-01

    The use of hydrogen-rich fuels may be challenging for burner designers due to unique properties of hydrogen compared to conventional fuels such as natural gas. Burner retrofit may be required to use hydrogen-enriched fuels in combustion systems that are designed for natural gas combustion. This study aimed to experimentally investigate NOx emissions from a novel low NOx burner fueled by methane-hydrogen mixtures. The burner was tested in a cylindrical combustion chamber at atmosph...

  20. Radiation exposure: Cytogenetic tests. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Forty test subjects who, either during or after the reactor accident of Chernobyl (26th April 1986), stayed at a building site at Shlobin 150 km away, were examined for spontaneously occurring as well as mitomycin C-induced Sister Chromatid Exchanges (SCE). The building site staff, who underwent a whole-body radionuclide count upon their return to Austria (June through September 1986), were used for the cytogenetic tests. The demonstration of the SCE was made from whole-blood cultures by the fluorescence/Giemse technique. At last 20 Metaphases of the 2nd mitotic cycle were evaluated per person. The radiation doses of the test subjects were calculated by adding the external exposure determined on the building site, the estimated thyroid dose through I-131, and the measured incorporation of Cs-134 and Cs-137. The subjects were divided into two groups for statistical analysis: One was a more exposed group (proven stay at Shlobin between 26th April and 31st May 1986, mostly working in the open air) and the other a less exposed group for comparison (staying at Shlobin from 1st Juni 1986 and working mainly indoors). (orig.)

  1. FIELD EVALUATION OF LOW-EMISSION COAL BURNER TECHNOLOGY ON UTILITY BOILERS. VOLUME I. DISTRIBUTED MIXING BURNER EVALUATION

    Science.gov (United States)

    The report gives results of a study in which NOx emissions and general combustion performance characteristics of four burners were evaluated under experimental furnace conditions. Of primary interest was the performance of a low NOx Distributed Mixing Burner (DMB), which was test...

  2. Ceramic application for regenerative burner system

    Energy Technology Data Exchange (ETDEWEB)

    Han, D.B.; Park, B.H.; Kim, Y.W.; Bae, W.S. [RIST, Pohang (Korea)

    1999-05-01

    Recently, regenerative burner system was developed and begins to be gradually used for better energy savings. Compared to conventional burner system, the regenerative one has the several merits such as higher fuel efficiency, light weight of apparatus, low harmful toxic gas and homogeneous heating zone, etc. The regenerative material, a very important component of the new regenerative burner system should possess the properties of low specific density, higher surface area and high specific heat capacity. Ceramics is the best regenerative material because of stable mechanical properties even at high temperature and better thermal properties and excellent chemical stability. In this study, alumina ball, alumina tube, 3-D ceramic foam and honeycomb as regenerative materials were tested and evaluated. The computer simulation was conducted and compared to the result of field test. This paper is aimed to introduce a new application of ceramics at high temperature. 7 refs., 5 figs., 3 tabs.

  3. SMART Reactor Flow Distribution Test (SCOP-E-01)

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. This report summaries and analyzes the SCOP-E-01 Test which simulated the reactor internal flow distribution under the steady state conditions with the same flow rate at each loop. The primary parameters, which are represented by static/differential pressure, flow rate and temperature were found to satisfy well the requirement of instrumentation and uncertainties. In order to evaluate overall quality of test results, various secondary parameters were selected and analyzed, which shows that the quality of data are good. From the various hydraulic data representing the hydraulics of the SMART reactor, the soundness and performance of the reactor design can be demonstrated. The test data will be utilized as boundary conditions for the thermal margin analysis of SMART reactor

  4. Downhole burner for wells

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, H.; Hazard, H.R.; Hummell, J.D.; Schulz, E.J.

    1966-03-22

    This is a downhole gas and air burner for use in wells to stimulate production. The combustible mixture is supplied to the combustion chamber of the downhole burner through a delivery tube. This tube includes a flow-back preventer and a check valve. The flashback preventers consist of a porous material which has restricted flow paths. The check valve controls the flow of combustible mixture to the combustion chamber and prevents undesirable pulsating flow through the combustion chamber and the delivery tube. The check valve also prevents flooding of the combustion chamber by well fluid. The burner is ignited electrically. The porous material can be flat strip or a conically shaped piece of thin porous metal.

  5. Reactor design of the SP-100 nuclear assembly test

    International Nuclear Information System (INIS)

    The Nuclear Assembly Test is currently being designed to demonstrate the performance characteristics of a 100-kWe version of the power source for the SP-100 Generic Flight System. Particular emphasis will be placed upon the operation of the prototypical ground test reactor under conditions of high-working temperatures and long life. The key features of the reactor include a small, compact core with component materials consisting of refractory metals and alloys. Because of the unique features of the SP-100 system, extensive use is made of Monte Carlo methods in the design and analysis of the reactor configuration. In addition, detailed testing of the reactor design has been carried out in the Zero Power Physics Reactor facility to provide calibration factors for the principal performance parameters. The key features of the test reactor design are described in this paper

  6. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1

  7. New Sensors for Irradiation Testing at Materials and Test Reactors

    International Nuclear Information System (INIS)

    Enhanced instrumentation, capable of providing real-time measurements of parameters during fuels and material irradiations, is required to support irradiation testing requested by US nuclear research programs. For example, several research programs funded by the US Department of Energy (US DOE) are emphasizing the use of first principle models to characterize the performance of fuels and materials. To facilitate this approach, high fidelity, real-time data are essential to demonstrate the performance of these new fuels and materials during irradiation testing. Furthermore, sensors that obtain such data in US MTRs, such as the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL), must be miniature, reliable, and able to withstand high fluxes and high temperatures. Depending on program requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these needs, INL has developed and deployed several new sensors to support irradiation testing in US DOE programs. The paper identifies the sensors currently available to support higher flux US MTR irradiations. Recent results and products from sensor research and development are highlighted. In particular, progress in deploying enhanced in-pile sensors for detecting temperature, elongation, and thermal conductivity is emphasized. Finally, initial results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are summarized. (author)

  8. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  9. Flat flame burner

    Energy Technology Data Exchange (ETDEWEB)

    Matsumura, Y.; Mitsudomi, H.

    1976-02-24

    Osaka Gas Co., Ltd.'s new flat-flame heat-treatment burner offers lower material costs, reduced combustion noise, and elimination of the need for a high-pressure fuel gas to provide a high-velocity combustion burner. The flat-flame burner contains an air-swirling chamber with a flame opening in one side; the wall defining the flame opening has a small thickness around the opening and a flat outer face. This construction causes the combustion gas to be forced out from the flame opening in a spiral direction by the swirling air current within the air chamber; together with the orifice effect of permitting the flame to emanate from a small opening to an unconfined outer space, this helps assure the formation of a flat flame spreading out over a very wide area for very rapid, uniform, and highly efficient heat treatment of an article to be heated. This approach also permits the thickness of the overall device to be reduced. The supply of combustion air in the form of a swirling stream makes it possible to provide a high-velocity combustion burner without using a high-pressure fuel gas, with the advantage of satisfactory mixture of the fuel gas and combustion air and consequently markedly reduced combustion noise.

  10. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  11. MINIMIZATION OF NO EMISSIONS FROM MULTI-BURNER COAL-FIRED BOILERS; SEMIANNUAL

    International Nuclear Information System (INIS)

    An initial testing campaign was carried out during the summer of 2000 to evaluate the impact of multiburner firing on NOx emissions. Extensive data had been collected during the Fall of 1999 and Spring of 2000 using a single pulverized-coal (PC) burner, and this data collection was funded by a separate Department of Energy program, the Combustion 2000 Low Emission Boiler System (LEBS) project under the direction of DB Riley. This single-burner data was thus available for comparison with NOx emissions obtained while firing three burners at the same overall load and operating conditions. A range of operating conditions were explored that were compatible with single-burner data, and thus the emission trends as a function of air staging, burner swirl and other parameters will be described below. In addition, a number of burner-to-burner operational variations were explored that provided interesing insight on their potential impact on NOx emissions. Some of these variations include: running one burner very fuel rich while running the others fuel lean; varying the swirl of a single burner while holding others constant; increasing the firing rate of a single burner while decreasing the others. In general, the results to date indicated that multiburner firing yielded higher NOx emissions than single burner firing at the same fuel rate and excess air. At very fuel rich burner stoichiometries (SR and lt; 0.75), the difference between multiple and single burners became indistinguishable. This result is consistent with previous single-burner data that showed that at very rich stoichiometries the NOx emissions became independent of burner settings such as air distributions, velocities and burner swirl

  12. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  13. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  14. Mobile reactor concepts as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ignition-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a four-rail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were unplugged and returned to a large, centrally located hot shop. A similar concept is envisioned for compact fusion reactor testing

  15. The mobile reactor concept as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ''ignition''-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a fourrail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were ''unplugged'' and returned to a large, centrally located ''hot'' shop. A similar concept is envisioned for compact fusion reactor testing

  16. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  17. Development, utilization, and future prospects of materials test reactors

    International Nuclear Information System (INIS)

    Reactor radiation affects the chemical and physical properties of materials. These changes can be very drastic in certain cases. Special test reactors have therefore been built since the 1950's and specific skills were developed to expose materials specimens to the precise irradiation conditions required. Materials testing reactors are those research reactor facilities which are designed and operated predominantly for studies into radiation damage. About a dozen plants in European communities (EC) Member States and in the US can be identified in this category, with 5 to 100 MW fission power and neutron fluxes between 5 x 1013 and 1015 cm-2s-1. The paper elaborates common aspects of development, utilization, and future prospects of US and EC materials testing reactors, and indicates the most significant differences

  18. THIRTY-DAY FIELD TESTS OF INDUSTRIAL BOILERS: SITE 5 - GAS-FIRED LOW-NOX BURNER

    Science.gov (United States)

    This is a final report on a test program to evaluate the long-term effectiveness of combustion modifications on industrial boilers. During previous programs, short-term tests were performed on industrial boilers to determine the effect of combustion modifications on air pollutant...

  19. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  20. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamic radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  1. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamics radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  2. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  3. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  4. Utilization of fission reactors for fusion engineering testing

    Energy Technology Data Exchange (ETDEWEB)

    Deis, G.A.; Miller, L.G.

    1985-02-08

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful.

  5. Recent reactor testing and experience with gamma thermometers

    International Nuclear Information System (INIS)

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith1. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here

  6. An improved method for fuel cycle analysis at equilibrium and its application to the study of fast burner reactors with variable conversion ratio

    International Nuclear Information System (INIS)

    Studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for detailed studies. In this paper we present an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the features of the fuel cycles parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different MA/Pu ratios in the fuel feed. It is shown that for cores with CR below ∼0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles. (authors)

  7. Reactor calculation benchmark PCA blind test results

    International Nuclear Information System (INIS)

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

  8. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  9. In-Research Reactor Tests for SCWR Fuel Verifications

    International Nuclear Information System (INIS)

    The Supercritical water cooled reactors (SCWRs) are essentially light water reactors (LWRs) operating at higher pressure and temperature. The SCWRs achieve high thermal efficiency (i.e., about 45% vs. about 35% efficiency for advanced LWRs) and are simpler plants as the need for many of the traditional LWR components is eliminated. The SCWRs build upon two proven technologies, the LWR and the supercritical coal-fired boiler. The main mission of the SCWR is production of low-cost electricity. Thus the SCWR is also suited for hydrogen generation with electrolysis, and can support the development of the hydrogen economy in the near term. In this paper, the SCWR fuel performance verification tests are reviewed. Based on this review results, in-research reactor verification tests to be performed in a fuel test loop through the international joint program are proposed. In addition, capsule tests and fuel test loop tests to be performed in HANARO are also proposed

  10. Micro-specimen testing techniques for evaluating nuclear reactor materials

    International Nuclear Information System (INIS)

    In the initial construction of nuclear power plant nuclear materials not only have to be high quality in mechanical properties and fracture resistant characteristics, but also considerations have to be given to weakness cause and continued safe operation of power reactor. Recognizing the importance of integrity evaluation test material samples are provided under monitoring program in reactor for evaluation of reactor material property. But because of limited space and necessity of a homogeneous irradiation environment a very limited quantity of micro specimen is provided. The existing test method of toughness property and fracture resistance requires pre-determined size specimen. Therefore, it is very difficult to evaluate those properties by limited micro-specimen provided under monitoring program. In this paper the test technologies of micro-specimen, which can be utilized to evaluate material integrity of reactors in operation, are reviewed. (Hong, J. S.)

  11. SMORN-III benchmark test on reactor noise analysis methods

    International Nuclear Information System (INIS)

    A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

  12. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  13. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  14. Dual-core TRIGA research and materials testing reactor

    International Nuclear Information System (INIS)

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  15. Results of the BREST-300 type reactor model fuel elements testing in the IGR reactor

    International Nuclear Information System (INIS)

    Testings of BREST-300 type fast reactor's model fuel elements with nitride fuel in the lead coolant in the central experimental channel of IGR reactor were carried out. In the testing the regime of non-controlled power burst was simulated. In the result of testing the seal failure of fuel elements with 2 % and 10 % 235U enrichment has been occurred, and fragmentation of the part of fuel pellets at interaction with coolant has been taken place. During the reactor testing the measurements and registration of experimental parameters (temperature of fuel, shell, coolant; pressure in fuel elements and testing ampoule; power release in the reactor) were conducted. The physical study of the 'fuel element - ampoule - reactor' was carried out, after-start-up spectrometric and material testing studies, calculated evaluation of temperature fields parameters in the testing ampoule were examined as well. Calculated and experimental values of breaking down specific power releases in the fuel are obtained. The assessment of both fuel fragmentation rate and it character is carried out. Distribution of fuel fragmentation within experimental ampoule volume is studied

  16. Oil burner nozzle

    Science.gov (United States)

    Wright, Donald G.

    1982-01-01

    An oil burner nozzle for use with liquid fuels and solid-containing liquid fuels. The nozzle comprises a fuel-carrying pipe, a barrel concentrically disposed about the pipe, and an outer sleeve retaining member for the barrel. An atomizing vapor passes along an axial passageway in the barrel, through a bore in the barrel and then along the outer surface of the front portion of the barrel. The atomizing vapor is directed by the outer sleeve across the path of the fuel as it emerges from the barrel. The fuel is atomized and may then be ignited.

  17. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional testing

    International Nuclear Information System (INIS)

    This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility. A vessel, designated ZB-1, has been tested under fatigue loadings with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialpruefungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE. AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions. (author)

  18. Technology Options for a Fast Spectrum Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  19. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

  20. Architecture of the ETR [experimental test reactor] systems code

    International Nuclear Information System (INIS)

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs

  1. P and T: An option for spent fuel and waste management using a double strata fuel cycle with a dedicated waste burner reactor

    International Nuclear Information System (INIS)

    The present commercial reactors (LWR, CANDU, etc.) operate in a Once Through Fuel Cycle OTC, and based in a feed of uranium. From around 400 operating reactors a large stock pile of radioactive waste are being produced, mainly long lived TRU- Plutonium, MA( Am, Np, Cm), and Long Lived Fission Products, LLFP, such as I-129, Tc- 99, Cs-135 etc. It is estimated around 300,000 t of the spent fuel be produced in this decade, with 1% of Pu (3,000 tons), 0.1% MA, 300 t, and 400 tons of LLFP. The build up of radioactive stock piles, besides the concern of waste disposal, also brings the issue of proliferation. To overcome these issues, the next generations of nuclear reactors are considering concepts that coupled with a closed fuel cycles in many new initiatives, such as GIF and INPRO. This is the main point to note, that is P and T is sustainable option for spent fuel and HLW management, considering the renascence of Nuclear Energy for the next decades. Some issues such as safety, economics had already been almost solved. The contribution of nuclear energy to avoid the threat of global warming due to CO2 emissions in short term is also a positive point. So the only point which still remains as a controversy issue for a complete acceptance of Nuclear Energy, is what is going to be done with the HLW (long term hazard). There is a need to give answers acceptable for the public, and as established in the Joint Convention for Safety Spent Fuel Management and Radioactive Waste Management to protect the people, the society, and the environment presently and in the future in such way that the needs from present generation be satisfied without compromising the future needs of the future generations. The scheme shown in this presentation, summarizes almost all possibilities of waste and spent fuel management. At present OTC cycle, only uranium is being used as fuel. The first point is that the utilization of thorium fuel cycle is an option to reduce long lived radio toxicity and

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  4. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  5. Models for transient analyses in advanced test reactors

    OpenAIRE

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world’s increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe whi...

  6. Testing of rubber O-rings for R-5 reactor

    International Nuclear Information System (INIS)

    This paper summarises the results of various experiments and tests conducted for the selection of suitable O-rings to be used in R-5 reactor. O-rings of various elastomeric compositions obtained from different manufacturers were tested for heat aging, fluid aging, radiation stability and specific gravity. They were irradiated to various dose levels at Apsara Reactor. The changes in axial thickness and hardness, after each test, were measured and results were correlated. The tests reveal that O-rings made of ethylene-propylene rubber are the best suited for the use in R-5 reactor. The O-rings made of nitrile rubber are also good. Neoprene rubber O-rings were found unsuitable mainly because of their low radiation resistance. (author)

  7. Monitoring test piece for a reactor pressure vessel

    International Nuclear Information System (INIS)

    Purpose: To obtain a test piece capable of measurement for neutron exposure ranging 0.1 -- 2 MeV in a reactor pressure vessel Constitution: Fissionable materials causing nuclear fission by fast neutrons are contained within a sealed container in addition to a test piece for monitoring the change in the mechanical properties and a monitor wire for measuring the neutron dose. If uranium 238 and thorium 232 are selected as the fissionable materials, for instance, they cause nuclear fission by the reaction with neutrons of higher than about 2 MeV and about 0.2 MeV respectively. Then, after the stop of the reactor operation, the monitoring test piece is taken out from the reactor pressure vessel to determine the radioactivity, whereby the neutron dose within the energy range of 0.1 - 2 MeV applied to the fissionable materials of the test piece can be estimated with ease. (Horiuchi, T.)

  8. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  9. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  10. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  11. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  12. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  13. Radial lean direct injection burner

    Science.gov (United States)

    Khan, Abdul Rafey; Kraemer, Gilbert Otto; Stevenson, Christian Xavier

    2012-09-04

    A burner for use in a gas turbine engine includes a burner tube having an inlet end and an outlet end; a plurality of air passages extending axially in the burner tube configured to convey air flows from the inlet end to the outlet end; a plurality of fuel passages extending axially along the burner tube and spaced around the plurality of air passage configured to convey fuel from the inlet end to the outlet end; and a radial air swirler provided at the outlet end configured to direct the air flows radially toward the outlet end and impart swirl to the air flows. The radial air swirler includes a plurality of vanes to direct and swirl the air flows and an end plate. The end plate includes a plurality of fuel injection holes to inject the fuel radially into the swirling air flows. A method of mixing air and fuel in a burner of a gas turbine is also provided. The burner includes a burner tube including an inlet end, an outlet end, a plurality of axial air passages, and a plurality of axial fuel passages. The method includes introducing an air flow into the air passages at the inlet end; introducing a fuel into fuel passages; swirling the air flow at the outlet end; and radially injecting the fuel into the swirling air flow.

  14. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  15. EMERIS: an advanced information system for a materials testing reactor

    International Nuclear Information System (INIS)

    The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

  16. Reactor Physics Tests for the Full Power Operation of HANARO

    International Nuclear Information System (INIS)

    The initial criticality of HANARO was achieved on the Feb. 8th of 1995. As HANARO is a unique reactor, there were difficulties to get a license to its full power operation, in which the design power of HANARO is 30 MW. There were two operation license conditions that limited the operation power to 80% of the design power. They were resolved in 2003 and the power ascension tests were conducted for the full power operation. This paper presents the several reactor physics tests for the power ascension to the full power of HANARO

  17. Low NO sub x regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Hovis, J.E.; Finke, H.P.

    1991-01-08

    This patent describes improvements in a regenerative burner having a regenerative bed, a burner port and a fuel nozzle. The improvement comprises: a burner baffle having apertures therein for selectively directing combustion air and inducing combustion gas recirculation into a primary combustion zone for suppressing NO{sub x} emissions, the baffle and the fuel nozzle being positioned substantially adjacent the burner port and being substantially coplanar in a plane perpendicular to a burner axis.

  18. Flat flame burner

    Energy Technology Data Exchange (ETDEWEB)

    Matsumura, Y.; Mitsudomi, H.

    1976-03-09

    Osaka Gas Co., Ltd.'s new flat-flame burner has an air-swirling chamber with a flame opening in one side so constructed that combustion gas is forced out from the flame opening in a spiral direction by the swirling air current within the air chamber. The orifice effect of permitting the flame to emanate from a small opening to an unconfined outer space assures formation of a flat flame spreading out over a very wide area, thereby ensuring very rapid, uniform and highly efficient heat treatment of an article to be heated. With the present invention, moreover, it is possible to materially reduce the thickness of the overall device.

  19. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  20. The RES Reactor. A test reactor for the French naval propulsion

    International Nuclear Information System (INIS)

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  1. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  2. REX 2000 core : a new material testing reactor project

    International Nuclear Information System (INIS)

    REX 2000 is a new research reactor project entirely dedicated to technological irradiations, which should be located on the CEA site of CADARACHE. It will be aimed at satisfying the future needs for the validation of new concepts of nuclear materials and fuels, and will take over and replace the present experimental reactors, which are 30 to 40 years old. The fundamental studies started by the CEA in 1993, on future irradiation needs expected in 2005, lead to the design of a reactor which will essentially meet the needs of PWRs, without forgetting the other fields such as FBRs, fusion... The current reactor project is based on a light water open pool concept, with a thermal power of 100 MW, in about 150 l, and characterized by an in-core-central hole. It reaches neutronic flux levels twice those of present French reactor fluxes. It allows many irradiations in the central loop under high fast neutron flux, in order to accelerate the aging of materials and analyze their behaviour. It also enables the achievement of power transient tests under high thermal neutron flux gradients. These performances are obtained with high forced flow rates and upward flow in the core, in order to preserve the operating flexibility of the reactor. This leads to the design of a specific assembly design. (author)

  3. PROTOTYPE EVALUATION OF COMMERCIAL SECOND GENERATION LOW-NO BURNER PERFORMANCE AND SULFUR CAPTURE

    Science.gov (United States)

    The report gives results of pilot-scale combustion tests of a Riley Stoker second-generation low-NOx burner combined with dry sorbent injection for SO2 control. The burner design is based on the distributed mixing concept. Combustion tests were conducted at 100 million Btu/hr in ...

  4. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235U loading in the reduced-enrichment fuel elements be the same as the 235U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant performance

  5. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  6. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  7. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  8. The technology development for surveillance test of reactor vessel materials

    International Nuclear Information System (INIS)

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs

  9. In situ tests on the PEC fast reactor building

    International Nuclear Information System (INIS)

    This paper describes forced excitation tests carried out at the PEC reactor building, to determine seismic motion amplifications produced in the building itself. Experimental results are used to gauge numerical methodologies capable of assessing the margins existing in the design analysis. (orig./HP)

  10. Natural convection test in Phenix reactor and associated CATHARE calculation

    International Nuclear Information System (INIS)

    The Phenix sodium cooled fast reactor started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, final tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test in Phenix reactor is the qualification of plant dynamic codes as CATHARE code for future safety studies. The paper firstly describes the Phenix reactor primary circuit. The initial test conditions and the detailed transient scenario are presented. Then, the CATHARE modelling of the Phenix primary circuit is described. The whole transient scenario is calculated, including the nominal state, the steam generators dry out, the scram, the onset of natural convection in the primary circuit and the natural convection phases. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram. Then, the evolution of main components inlet and outlet temperatures is compared. The need of coupling a system code with a CFD code to model the 3D behaviour of large pools is pointed out. This work is in progress. (author)

  11. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U3Si2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  12. Sipping test of fuel assemblies in LVR-15 reactor

    International Nuclear Information System (INIS)

    The LVR-15 reactor is a light water research type which is situated at NRI in Rez near Prague. The poster describes the procedure and methodology used for sipping test of the fuel assemblies. These tests are designed to evaluate the leakage of fuel and fission products from the tested fuel assembly. From 1995 to 2003 there have been performed about 200 tests. Examples of results of sipping water activity measurements are presented. The values of activities of 137Cs and 134Cs are used for decision if the fuel assembly can be used in reactor core, transported to storage pool or if it is necessary to put the fuel assembly into the special protective can. The used limits of activities are discussed. (author)

  13. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium (3. However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

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

    International Nuclear Information System (INIS)

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

  15. Integrated leak rate test results of JOYO reactor containment vessel

    International Nuclear Information System (INIS)

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

  16. Fusion-Fission Burner for Transuranic Actinides

    Science.gov (United States)

    Choi, Chan

    2013-10-01

    The 14-MeV DT fusion neutron spectrum from mirror confinement fusion can provide a unique capability to transmute the transuranic isotopes from light water reactors (LWR). The transuranic (TRU) actinides, high-level radioactive wastes, from spent LWR fuel pose serious worldwide problem with long-term decay heat and radiotoxicity. However, ``transmuted'' TRU actinides can not only reduce the inventory of the TRU in the spent fuel repository but also generate additional energy. Typical commercial LWR fuel assemblies for BWR (boiling water reactor) and PWR (pressurized water reactor) measure its assembly lengths with 4.470 m and 4.059 m, respectively, while its corresponding fuel rod lengths are 4.064 m and 3.851 m. Mirror-based fusion reactor has inherently simple geometry for transmutation blanket with steady-state reactor operation. Recent development of gas-dynamic mirror configuration has additional attractive feature with reduced size in central plasma chamber, thus providing a unique capability for incorporating the spent fuel assemblies into transmutation blanket designs. The system parameters for the gas-dynamic mirror-based hybrid burner will be discussed.

  17. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAlx) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( ∼ 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient calculations

  18. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm3 was by then in routine use, illustrated how far work has progressed

  19. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

  20. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  1. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  2. Reactor vessel model flow tests for 145-fuel assembly core

    International Nuclear Information System (INIS)

    Hydraulic tests on a one-sixth-scale model of a two-loop pressurized water reactor with 145 fuel assemblies are described. Core inlet and outlet flow distributions and reactor vessel pressure drop were investigated. The core inlet flow distribution was developed to be independent of the flow conditions in the inlet annulus. A flow distribution system, consisting of several flow splitters in the inlet annulus and a spherical plate flow distributor in the lower head region, was developed to obtain a symmetric and stable core inlet flow distribution. A minimum core inlet flow factor of 0.99 was established in the core. Reactor vessel unrecoverable pressure drops were measured on the model to predict losses that will occur in the prototype

  3. Research on reactor physics using the Japan Materials Testing Reactor Critical Facility (JMTRC)

    International Nuclear Information System (INIS)

    The JMTRC of 100 W was installed for the purpose of carrying out the basic experiment on the nuclear characteristics of reactors and the preceding test related to the operation plan of the Japan material testing reactor (JMTR, 50 MW). After the attainment of the initial criticality in October, 1965, for obtaining the reactor physics characteristics, criticality experiment was begun. The items of the criticality experiment were critical mass, control rod worth, reactor dynamic characteristic parameters, shutdown margin and so on, and these experimental data were effectively utilized for the safety evaluation in the operation of the JMTR. The preceding test using the JMTRC has been carried out for obtaining the nuclear characteristics of samples and the thermal characteristics estimated from those results by simulating the JMTR core. In August, 1983, the degree of fuel enrichment for the JMTRC was reduced to 45 % U-235, and various experiments usig the MEU core were carried out. In this paper, the criticality experiment using the MEU core and the experiment on the characteristics of lithium-containing pellets are reported. (K.I.)

  4. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  5. Experience with the generating plant at fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWth/13.2 MW(e) sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and generating systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct.1985 with Mark-I core (70 % PUC-30 % UC). FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. Heat generated in the reactor core is transported to the tertiary loop by primary and secondary sodium loops. The steam water system mainly consists of a once through steam generator, which produces super heated steam at a pressure of 120 bars and temperature of 480 degC, feed water system and condensate system. The steam produced is supplied to a condensing turbine. The turbine in turn is coupled to an alternator. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. The paper presents operating experience with generating plant consisting of steam water circuit, condensing turbine and its associated systems and the alternator, various modifications carried out to improve system reliability and availability and certain incidents taken place in the generating plant. (author)

  6. The decommissioning of the KEMA suspension test reactor

    International Nuclear Information System (INIS)

    In this report the decommissioning of the KEMA Suspension Test Reactor (KSTR) is described. This reactor was a 1 MWth aqueous homo-geneous nuclear reactor in which a suspension of a mixed oxide UO2/ ThO2 in light water was circulated in a closed loop through a sphere-shaped core vessel. The reactor, located on KEMA premises, made 150 MW of heat during its critical periods. Dismantling of this reactor, with its many connected subsystems, meant the mastering of activated components which were also contaminated on inner surfaces caused by small fuel deposits (alpha contaminants) and fission products (beta, gamma contaminants). A description is given of the save removal of the fuel, the remote dismantling of systems and components and the disposal of steel scrap and other materials. Important features are the measures to be taken and provisions needed for safe handling, for the reduction of the radiation dose for the working team and the prevention of spreading of activity over the working area and the environment. It has been demonstrated that safe dismantling and disposal of such systems can be achieved. Experience gained at KEMA for the proper dismantling and for safety measures to be taken for workers and the environment can be made available for similar dismantling projects. A cost break-down is included in the report. (author). 22 refs.; 52 figs.; 12 tabs

  7. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.)

  8. Computational investigations of low-emission burner facilities for char gas burning in a power boiler

    Science.gov (United States)

    Roslyakov, P. V.; Morozov, I. V.; Zaychenko, M. N.; Sidorkin, V. T.

    2016-04-01

    Various variants for the structure of low-emission burner facilities, which are meant for char gas burning in an operating TP-101 boiler of the Estonia power plant, are considered. The planned increase in volumes of shale reprocessing and, correspondingly, a rise in char gas volumes cause the necessity in their cocombustion. In this connection, there was a need to develop a burner facility with a given capacity, which yields effective char gas burning with the fulfillment of reliability and environmental requirements. For this purpose, the burner structure base was based on the staging burning of fuel with the gas recirculation. As a result of the preliminary analysis of possible structure variants, three types of early well-operated burner facilities were chosen: vortex burner with the supply of recirculation gases into the secondary air, vortex burner with the baffle supply of recirculation gases between flows of the primary and secondary air, and burner facility with the vortex pilot burner. Optimum structural characteristics and operation parameters were determined using numerical experiments. These experiments using ANSYS CFX bundled software of computational hydrodynamics were carried out with simulation of mixing, ignition, and burning of char gas. Numerical experiments determined the structural and operation parameters, which gave effective char gas burning and corresponded to required environmental standard on nitrogen oxide emission, for every type of the burner facility. The burner facility for char gas burning with the pilot diffusion burner in the central part was developed and made subject to computation results. Preliminary verification nature tests on the TP-101 boiler showed that the actual content of nitrogen oxides in burner flames of char gas did not exceed a claimed concentration of 150 ppm (200 mg/m3).

  9. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  10. Verification tests performed for development of an integral type reactor

    International Nuclear Information System (INIS)

    SMART is an integral type reactor with innovative design features aimed at achieving a highly enhanced safety and improved economics. The SMART design is based on proven reactor design technologies with the use of new advanced design features. Most of the design features implemented into the SMART have been proven, however the advanced design features implemented into the SMART should be proven by testing. Various thermal hydraulic experiments have been carried out and also planned to assure the fundamental behavior of major concepts of the SMART and to prove the performance of the systems with new innovative technologies. This paper describes the thermal hydraulic test program for the SMART development and briefly discusses the typical test results. (author)

  11. Test reactor studies of the shadow corrosion phenomenon

    International Nuclear Information System (INIS)

    The shadow effect, resulting in enhanced corrosion on zirconium-base alloys in proximity to other metal, has been observed since the 1960's. In 1997, hot-cell examinations revealed thick oxide layers on fuel cladding surfaces that during irradiation had been located in the shadow of Inconel spacer grids. The acronym ESSC (Enhanced spacer shadow corrosion) was used to describe the phenomenon. Following the observation of ESSC, several investigations were initiated to develop an explanation of the shadow corrosion phenomenon. In this paper, results are presented and compared from in-reactor experiments performed in three different test reactors, the R2 reactor in Studsvik, the MITR-II reactor at MIT, and the Halden test reactor. The first study, performed in Studsvik, was initiated to assess the feasibility of studying the shadow corrosion phenomenon during relatively short irradiation periods in a test reactor. The results clearly showed that a shadow was formed on Zircaloy cladding in contact with Inconel within 34 days of exposure in the center of the core, while no shadow was observed on specimens situated outside and upstream from the core. The test at MIT was initiated to identify the basic mechanisms of the shadow effect. The MIT research reactor MITR-II was used to simulate BWR core coolant conditions. A sample train included Zircaloy-2 claddings with different counter materials surrounding each clad specimen. The counter materials were intended to serve as the materials tentatively producing shadow corrosion. when located in close proximity to the clad specimens. The counter materials chosen were strong or weak beta emitters, platinum, inert material (zirconia), Inconel X-750, coated Inconel X-750, and Zircaloy-2 at various separation distances. The results showed that beta radiation from the counter material is not the main mechanism for the shadow effect, that a coating on the counter material inhibits the formation of shadow corrosion, and that shadow

  12. Test reactor studies of the shadow corrosion phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, B.; Limback, M. [Westinghouse Atom AB, Vasteras (Sweden); Wikmark, G. [Advanced Nuclear Technology Swede AB, Uppsala (Sweden); Hauso, E.; Johnsen, T. [Institut fur Energiteknikk, Halden (Norway); Ballinger, R.G. [Massachusetts Institute of Technology, Cambridge, Ma (United States); Nystrand, A.C. [Studsvik Nuclear AB, Nykiping (Sweden)

    2002-07-01

    The shadow effect, resulting in enhanced corrosion on zirconium-base alloys in proximity to other metal, has been observed since the 1960's. In 1997, hot-cell examinations revealed thick oxide layers on fuel cladding surfaces that during irradiation had been located in the shadow of Inconel spacer grids. The acronym ESSC (Enhanced spacer shadow corrosion) was used to describe the phenomenon. Following the observation of ESSC, several investigations were initiated to develop an explanation of the shadow corrosion phenomenon. In this paper, results are presented and compared from in-reactor experiments performed in three different test reactors, the R2 reactor in Studsvik, the MITR-II reactor at MIT, and the Halden test reactor. The first study, performed in Studsvik, was initiated to assess the feasibility of studying the shadow corrosion phenomenon during relatively short irradiation periods in a test reactor. The results clearly showed that a shadow was formed on Zircaloy cladding in contact with Inconel within 34 days of exposure in the center of the core, while no shadow was observed on specimens situated outside and upstream from the core. The test at MIT was initiated to identify the basic mechanisms of the shadow effect. The MIT research reactor MITR-II was used to simulate BWR core coolant conditions. A sample train included Zircaloy-2 claddings with different counter materials surrounding each clad specimen. The counter materials were intended to serve as the materials tentatively producing shadow corrosion. when located in close proximity to the clad specimens. The counter materials chosen were strong or weak beta emitters, platinum, inert material (zirconia), Inconel X-750, coated Inconel X-750, and Zircaloy-2 at various separation distances. The results showed that beta radiation from the counter material is not the main mechanism for the shadow effect, that a coating on the counter material inhibits the formation of shadow corrosion, and that

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

    International Nuclear Information System (INIS)

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

  14. Enhanced in-pile instrumentation at the advanced test reactor

    International Nuclear Information System (INIS)

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  15. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  16. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  17. Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

    International Nuclear Information System (INIS)

    The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

  18. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4∼2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure

  19. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  20. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  1. The Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

    2011-08-01

    In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are

  2. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  3. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  4. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  5. Oil fired boiler/solar tank- and natural gas burner/solar tank-units

    DEFF Research Database (Denmark)

    Furbo, Simon; Vejen, Niels Kristian; Frederiksen, Karsten Vinkler

    1999-01-01

    During the last few years new units consisting of a solar tank and either an oil fired boiler or a natural gas burner have been introduced on the Danish market. Three different marketed units - two based on a natural gas burner and one based on an oil fired boiler - have been tested in a heat...

  6. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  7. New results from pulse tests in the CABRI reactor

    International Nuclear Information System (INIS)

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared

  8. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  9. Potential for new societal contributions from the advanced test reactor

    International Nuclear Information System (INIS)

    The mission of the Advanced Test Reactor (ATR) at Idaho National Engineering Laboratory is to study the effects of intense radiation on materials and fuels and to produce radioisotopes for the U.S. Department of Energy (DOE) for government and commercial applications. The purpose of this paper is to explore the potential benefits to society from these available neutrons. The ATR is a 250-MW(thermal) light water reactor highly enriched uranium in plate-type fuel. The ATR uses a combination of hafnium control drums and shim rods to adjust power and hold flux distortion to a minimum. The different quadrants of the ATR can be operated at different power levels to meet a variety of mission requirements. Irradiation positions are available at various locations throughout the core and beryllium reflector. A summary of the flux levels at various ATR reflector and loop positions. is given. These fluxes are maintained with a relatively constant axial flux profile throughout cycles that last 35 to 42 days. These neutrons can be used for testing and irradiation programs that support commercial reactor license extension, advanced fuel development, materials effects studies, failure cause/effect studies, coolant chemistry evaluations, prototype testing programs, isotope production, and basic research. Radioisotope production falls into three categories: medical, industrial, and research. In summary, the ATR is a unique, high-power test reactor capable of supporting the current DOE mission and producing radioisotopes. Space available for radioisotope production and fuels or materials testing will increase by 44% in 1994, improving DOE's ability to support national needs in health care, industry, and research

  10. Reactor vessel integrity analysis based upon large scale test results

    International Nuclear Information System (INIS)

    The fracture mechanics analysis of a nuclear reactor pressure vessel is discussed to illustrate the impact of knowledge gained by large scale testing on the demonstration of the integrity of such a vessel. The analysis must be able to predict crack initiation, arrest and reinitiation. The basis for the capability to make each prediction, including the large scale test information which is judged appropriate, is identified and the confidence in the applicability of the experimental data to a vessel is discussed. Where there is inadequate data to make a prediction with confidence or where there are apparently conflicting data, recommendations for future testing are presented. 15 refs., 6 figs.. 1 tab

  11. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of

  12. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  13. Flow processes in a radiant tube burner: Combusting flow

    International Nuclear Information System (INIS)

    Highlights: → 3D combusting flow in an industrial radiant tube burner is modelled using the ANSYS-CFX CFD code. → Results are validated against data from an industrial furnace (NO emissions within 7%). → The flame is long and narrow with slight asymmetry. Mixing near the fuel injector is very effective. → The recuperator section is reasonably effective, but design improvements are proposed. → The design is vulnerable to eccentricities due to manufacturing or assembly tolerances. -- Abstract: This paper describes a study of the combustion process in an industrial radiant tube burner (RTB), used in heat treating furnaces, as part of an attempt to improve burner performance. A detailed three-dimensional Computational Fluid Dynamics model has been used, validated with experimental test furnace temperature and flue gas composition measurements. Simulations using the Eddy Dissipation combustion model with peak temperature limitation and the Discrete Transfer radiation model showed good agreement with temperature measurements in the inner and outer walls of the burner, as well as with flue gas composition measured at the exhaust (including NO). Other combustion and radiation models were also tested but gave inferior results in various aspects. The effects of certain RTB design features are analysed, and an analysis of the heat transfer processes within the burner is presented.

  14. Fuels for research and test reactors, status review: July 1982

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO/sub 2/ rod fuels. Among new fuels, those given major emphasis include H/sub 3/Si-Al dispersion and UO/sub 2/ caramel plate fuels.

  15. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO2 rod fuels. Among new fuels, those given major emphasis include H3Si-Al dispersion and UO2 caramel plate fuels

  16. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  17. Is it possible to test the LSND parameters at reactors?

    International Nuclear Information System (INIS)

    This report describes a part of the programme that was reported at the conference NANPino 2000. We propose a two-detector experiment at Krasnoyarsk for studying small mixing parameters at atmospheric neutrino mass range. As a first stage we consider here an experiment with two much smaller detectors that can test the LSND region. In the framework of our programme we also study the reactor antineutrino energy spectrum

  18. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  19. Testing Of Secondary Cooling Component Of TRIGA Mark Reactor

    International Nuclear Information System (INIS)

    The aim of this activity is to improve the knowledge of the mechanical testing technology of the research reactor cooling pipe material. The way which was chosen is through a series of testing to know the mechanical properties of carbon steel pipe used in TRIGA-MARK II secondary cooling pipe. Scopes of these testing activities are tensile testing, hardness testing, chemical composition analysis, and metallography analysis. Visual examination shows that thickness of the pipe was reduced over the range 0.31-1.76 mm and there was scales inside the pipe about 7.1-9.1 mm. Result of the mechanical testing shows that ultimate tensile strength, yield strength, elongation and. hardness of that material are 39 kg mm2, 34 kg/mm2, 38 %, and HV161, respectively. That yield strength value is on the design range

  20. Acceptance test of graphite components in nuclear reactor

    International Nuclear Information System (INIS)

    The HTTR is the first high temperature gas-cooled reactor in Japan. It is a test reactor with thermal power of 30 MW and coolant outlet temperature of 950degC at maximum. To achieve high temperature coolant core internals were made of graphite and carbon materials due to their excellent thermal resistivity. After fabrication of graphite and carbon components at works they were installed in the HTTR, and now it is in the power up testing stage. Concerning the inspection standard of the graphite and carbon components, nondomestic standard exists as main components in the nuclear reactor. It is necessary, therefore, to prescribe the inspection standards for the HTTR graphite components. Many research and developments in relation to the inspection standard, e.g. in the research field of nondestructive examination of the graphite material, had been performed, and then the JAERI established the inspection standard. The acceptance test of the graphite and carbon components was carried out based on the inspection standard. This paper prescribes the outline of the established inspection standard. (author)

  1. Regenerative ceramic burner has highest efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gettings, M.

    1986-01-01

    Regenerative ceramic burners consisting of a double gas/air burner and utilising waste heat which is stored via regenerators are described. The system is capable of operating at 1400/sup 0/C, it removes about 85-90% of energy from hot waste gases and exhibits energy savings of 40-60% over cold nozzle mix burners and 20-25% over recuperative burners. (UK).

  2. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    International Nuclear Information System (INIS)

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed (Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008). This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  3. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  4. A gas burner device for highspeed heating

    Energy Technology Data Exchange (ETDEWEB)

    Nosach, V.G.; Danchenko, V.N.; Zanemonets, V.F.

    1979-01-01

    The design and the results of the investigations of gas burners with porous cooling by fire Pv of two forms: a gas burner which makes it possible to organize the process of the burning for Pv heating, and a gas burner creating a stream of combustion products.

  5. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements

  6. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  7. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... 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)....

  8. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

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

    International Nuclear Information System (INIS)

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

  10. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  11. Temperature controlled material irradiation in the advanced test reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor's principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor's capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment 'in-pile tube (IPT)' inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities

  12. CANDU load following test in ICN research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbas, S.; Palleck, S. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Hohoianu, G.; Parvan, M. [Inst. for Nuclear Research, ICN (Romania)

    2008-07-01

    To study the performance of CANDU fuel under Load Following (LF) conditions, a CANDU 6 type fuel element with 8.0 wt% {sup 235}U enrichment was irradiated in the 14 MW TRIGA materials testing reactor at the Institute for Nuclear Research (ICN) in Romania. This experiment was developed under the INR-AECL Memorandum for Co-operation in research and development of nuclear energy and technology. The fuel element underwent a successful demonstration of LF capability, where the fuel element withstood 200 daily cycles from 27 to 54 kW/m (average element linear power), as well as additional ramps due to reactor trips and restarts during the test period. The fuel element underwent a series of post-irradiation destructive and nondestructive examinations after the LF test irradiation. No performance or integrity issues were observed. This paper presents a description of the test facility, details of the test irradiation conditions and the post-irradiation examination results with discussion on their relation to CANDU fuel performance and integrity. (author)

  13. Design of CAP1400 reactor internals flow-induced vibration simulation test model

    International Nuclear Information System (INIS)

    While the first CAP1400 reactor internals is defined as a 'prototype', it is necessary to carry out the reactor internals flow-induced vibration simulation test for verifying the structure integrity of reactor internals and providing data for vibration assessment of CAP1400 reactor internals. The reactor internals flow-induced vibration simulation test is usually a reduced scale model test. This paper describes in detail the main factors of model scale, similarity criterion and the simplification of test model. The simplification of model will not only reduce the fabrication cost, but also obtain more accurate test data from the flow-induced vibration simulation test. (authors)

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

    International Nuclear Information System (INIS)

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

  15. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor

  16. A linear model of the Fast Breeder Test Reactor Plant

    International Nuclear Information System (INIS)

    A linear analysis of the Fast Breeder Test Reactor System, consisting of the reactor, intermediate heat exchanger, steam generator and connected piping is presented. The problem of variable boundaries in the steam generator is reduced to a problem of fixed boundaries by dividing the steam generator into six zones. Based upon this, one can obtain the transfer function of any input/output combination. Starting with the time domain non-linear partial differential equations, the problem is reduced to a system of linear equations in complex variables, which can be solved basically by Gaussian elimination process. The results of this work will be useful in determining a suitable control scheme for waterflow in the steam generator and the control parameters. (auth.)

  17. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  18. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    International Nuclear Information System (INIS)

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm·m3/m·m), capacity; 3115m3/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm·m3/m·m), capacity; 539.4m3/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m3/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m3/h, total head; 42.0m, motor output; 190kw

  19. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B.

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black �control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized

  20. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  1. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  2. ENVIRONMENTAL ASSESSMENT OF AN ENHANCED OIL RECOVERY STEAM GENERATOR EQUIPPED WITH A LOW-NOX BURNER. VOLUME 1. TECHNICAL RESULTS

    Science.gov (United States)

    The report discusses results from sampling flue gas from an enhanced oil recovery steam generator (EOR steamer) equipped with an MHI PM low-NOx burner. The tests included burner performance/emission mapping tests, comparative testing of an identical steamer equipped with a conven...

  3. High temperature indentation tests on fusion reactor candidate materials

    International Nuclear Information System (INIS)

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

  4. The Jules Horowitz Reactor (JHR), a European Material Testing Reactor (MTR), with extended experimental capabilities

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-1, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

  5. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-1, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  6. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional testing

    International Nuclear Information System (INIS)

    The objective of the acoustic emission (AE)/flaw characterization program is to provide an experimental feasibility evaluation of using the AE method on a continuous basis (during operation and during hydrotest) to detect and analyze flaw growth in reactor pressure vessels and primary piping. This effort is based on earlier results showing that AE has potential for being a valuable addition to nondestructive evaluation (NDE) methods with the added unique capability for continuous monitoring, high sensitivity and remote flaw location. Results are reported for the ZB-1 vessel test and the Watts Bar-1 hot functional test

  7. Local stability tests in Dresden 2 boiling water reactor

    International Nuclear Information System (INIS)

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations

  8. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, V.V.; Dubrovin, K.P.; Ivanov, E.G.; Korneev, V.T.; Kruglov, A.B.; Lebedev, L.M.

    1987-11-01

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < ..mu..m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).

  9. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    International Nuclear Information System (INIS)

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < μm thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level)

  10. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    International Nuclear Information System (INIS)

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author)

  11. 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.

    2006-07-01

    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

  12. Fracture toughness test methods and examples for fusion reactor materials

    International Nuclear Information System (INIS)

    This paper introduces the importance of the evaluation of fracture toughness in nuclear fusion reactor structural materials, and the fracture toughness evaluation methods that are used as the standards and their actual examples. It also discusses the problems involved in the standardized approach and the efforts for the technology improvement. To evaluate the material life under nuclear fusion reactor environment, fracture toughness measurement after neutron irradiation is indispensable. Due to a limitation in the irradiation area size of an irradiation reactor, and to avoid the temperature difference in a specimen, the size of the specimen is required to be minimized, which is different from the common standards. As for the size effect of the test specimen, toughness value tends to decrease when ligament length is 7 mm or below. The main problems and challenges are as follows. (1) As for the tendency that fracture toughness value decreases along with the miniaturization of the ligament length, it is necessary to elucidate the mechanism of size effects, and to develop the correction method for size effects. (2) As for the issues of the curve shape and application to irradiation time in the master curve method, it is necessary to review the data checking method and plastic constraint conditions for crack tip M = 30 that is stipulated in ASTM E1921, and to elucidate the material dependence of master curve shape. (A.O.)

  13. Status of research reactor fuel test in the High Flux Reactor (Petten)

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Casalta, S. [European Commission, JRC, NL-1755 ZG Petten (Netherlands); Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.; Dassei, G. [NRG Petten (Netherlands); Vacelet, H. [Cerca Framatome, F-26104 Romans (France); Languille, A. [CEA Cadarache, F-13108 Saint Paul Lez Durance (France)

    2001-07-01

    Even if the research reactors are using very well known MTR-fuel, a need exists for research in this field mainly for the reasons of industrial qualification of fuel assemblies (built with qualified fuel), improvement or modifications on a qualified fuel ( e.g. increase of density), and qualification of a new fuels such as UMo. For these types of tests, the High Flux Reactor located in Petten (the Netherlands) has a lot of specific advantages: 1) a large core with various interesting positions ranging from high to low fluence rate; 2) a high number of operating days (>280 days/year) that gives - with the high flux available - a possibility to reach quickly high burnup; 3) a downward coolant flow that simplifies the device engineering; 4) all possibilities of non-destructive and destructive examinations in the hot-cells (visual inspection, swelling, {gamma}-scanning, macro- and light microscopy, SEM and EPMA examinations, tomography). Two types of tests can be performed at the plant: either a full-scale test or a test of plates in dedicated devices. A presentation is made of the irradiation test on four UMo plates, begun in March 2000 in the device UMUS. A status report is provided of the full-scale test to be done in the near future, especially the UMo tests to begin the next year. In conclusion it appears that the HFR, that had already given an excellent contribution to silicide fuel qualification in the 1980s, will also give a significant contribution to the current UMo qualification programs. (author)

  14. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

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

    International Nuclear Information System (INIS)

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

  16. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  17. Experimental and numerical analysis of isothermal turbulent flows in interacting low NOx burners in coal-fired furnaces

    OpenAIRE

    Cvoro, Valentina

    2007-01-01

    Coal firing power stations represent the second largest source of global NOx emissions. The current practice of predicting likely exit NOx levels from multi-burner furnaces on the basis of single burner test rig data has been proven inadequate. Therefore, to further improve current NOx reduction technologies and assist in the assessment of NOx levels in new and retrofit plant cases, an improved understanding of the impact of burner interactions is required. The aim of this research is tw...

  18. Fast Shutdown System tests in the Georgia Tech Research Reactor

    International Nuclear Information System (INIS)

    The Fast Shutdown System (FSS) is a new safety system design concept being considered for in installation in the Savannah River (SRS) production reactors. This system is expected to mitigate the consequences of a Design Basis Loss of Coolant Accident, and therefore allow higher operational power levels. A test of this system in the Georgia Tech Research Reactor is proposed to demonstrate the efficacy of this concept. Three tests will be conducted at full power (5MW) and one at low power (100kw). Two full power tests will be conducted with the FSS rod backfilled with one (1) atmosphere of He-4, and one with the rod evacuated. The low power conducted with the FSS rod evacuated. Neutron flux and pressure data will be collected with an independent data acquisition system (DAS). Safety issues associated with the performance of the Fast Shutdown System experiments are addressed in this report. The credible accident scenarios were analyzed using worst case scenarios to demonstrate that no significant nuclear or personnel safety hazards would result from the performance of the proposed experiments

  19. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  20. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  1. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    OpenAIRE

    Wang, Guoliang; Jensen, Peter Arendt; Hao WU; Bøjer, Martin; Jappe Frandsen, Flemming; Glarborg, Peter

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured, forming water-insoluble K-aluminosilicate. The amount of K captured by 1 g kaolin rose when increasing the molar ratio of K/Si in the reactant. Changing of reaction temperature from 1100 °C to 1300 °C...

  2. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao;

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......, forming water-insoluble K-aluminosilicate. The amount of K captured by 1 g kaolin rose when increasing the molar ratio of K/Si in the reactant. Changing of reaction temperature from 1100 °C to 1300 °C did not influence the extent of reaction, which is different from the results observed in previous fixed...

  3. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately

  4. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  5. EPR/PTFE dosimetry for test reactor environments

    International Nuclear Information System (INIS)

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  6. Program plan for decontamination and decommissioning the Materials Testing Reactor at the INEL

    International Nuclear Information System (INIS)

    A discussion is presented of a program plan developed for the dismantling of the Materials Testing Reactor located in the Testing Reactor Area (TRA) of the Idaho National Engineering Laboratory. Included are the scope of work, dismantling problems resulting from the nature of construction of the MTR, and a program plan for physically dismantling the reactor

  7. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  8. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional

    International Nuclear Information System (INIS)

    The objective of the acoustic emission (AE)/flaw characterization program presented is to develop use of the AE method on a continuous basis during operation and during hydrotest, to detect and analyze flow growth in reactor pressure vessels and primary piping. The program scope is described by three primary areas of effort: develop a method to identify crack growth AE signals; develop a relationship between measured AE and crack growth; demonstrate the total concept through off-reactor vessel tests; and, on-reactor monitoring. The laboratory speciments used to determine fundamental feasibility of program objectives were ASTM A533 B, Class 1 steel. The ZB-1 vessel test is described, and the results are presented. Reactor hot functional testing was done on the Watts Bar Unit 1. Evidence shows that AE from cracking in inaccessible parts of the reactor system such as the vessel beltline should be detectable

  9. Performance analysis of the 840 MWt PRISM reference burner core

    International Nuclear Information System (INIS)

    The General Electric PRISM (Power Reactor, Innovative Small Module) is a modular, pool-type sodium-cooled fast reactor employing innovative, passive features to provide an extremely high level of public safety. A PRISM power block consists of two 840 MWt reactor modules, each with a vessel diameter of 9.15 m (30 ft), tied to a turbine generator and producing 622 MWe. A full-size plant consists of three power blocks producing 1866 MWe of electrical power. Two core configurations have been analyzed. The reference is a 'burner' core (conversion ratio of 0.8) and the alternative is a breakeven' core (plutonium consumption balanced by plutonium generation). The core nuclear designs are largely governed by passive safety and reactivity control issues. The key features employed to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters (rod stops) and gas expansion modules (GEMs). A passive reactor vessel auxiliary cooling system (RVACS) assures safety-grade decay heat removal. This paper summarizes the operational and safety performance of the 840 MWt PRISM modular reactor, with emphasis on the reference burner core. (author)

  10. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  11. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Science.gov (United States)

    2010-01-01

    ... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2010-01-01 2010-01-01 false Primary Reactor Containment Leakage Testing for Water... PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor...

  12. Action Memorandum for Decommissioning the Engineering Test Reactor Complex under the Idaho Cleanup Project

    International Nuclear Information System (INIS)

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared and released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessel. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface

  13. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  14. Test on the reactor with the intelligent extrapolation criticality device for physical startup experiment

    International Nuclear Information System (INIS)

    The Intelligent Extrapolation Criticality Device is used for automatic counting and automatic extrapolation during the criticality experiment on the reactor. Test must be performed on the zero-power reactor or other reactor before the Device is used. The paper describes the test situation and test results of the Device on the zero-power reactor. The test results show that the Device has the function of automatic counting and automatic extrapolation, the deviation of the extrapolation data is small, and it can satisfy the requirements of physical startup on the reactor. (author)

  15. Testing of the thermoemission reactor-converters Topaz-I and Topaz-II

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, V.A.

    1972-12-31

    The construction of reactor-converters and their neutronphysics characteristics are outlined. The test equipment is described. Problems encountered in measuring the electric power output of the reactor-conventer are discussed. The basic results obtained are summarized. (JSR)

  16. The Advanced Test Reactor as a National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) has been in operation since 1967 and mainly used to support U.S. Department of Energy (US DOE) materials and fuels research programs. Irradiation capabilities of the ATR and post-irradiation examination capabilities of the Idaho National Laboratory (INL) were generally not being utilized by universities and other potential users due largely to a prohibitive pricing structure. While materials and fuels testing programs using the ATR continue to be needed for US DOE programs such as the Advanced Fuel Cycle Initiative and Next Generation Nuclear Plant, US DOE recognized there was a national need to make these capabilities available to a broader user base. In April 2007, the U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). As a NSUF, most of the services associated with university experiment irradiation and post-irradiation examinations are provided free-of-charge. The US DOE is providing these services to support U.S. leadership in nuclear science, technology, and education and to encourage active university/industry/laboratory collaboration. The first full year of implementing the user facility concept was 2008 and it was a very successful year. The first university experiment pilot project was developed in collaboration with the University of Wisconsin and began irradiation in the ATR in 2008. Lessons learned from this pilot program will be applied to future NSUF projects. Five other university experiments were also competitively selected in March 2008 from the initial solicitation for proposals. The NSUF now has a continually open process where universities can submit proposals as they are ready. Plans are to invest in new and upgraded capabilities at the ATR, post-irradiation examination capabilities at the INL, and in a new experiment assembly facility to further support the implementation of the user facility concept. Through a newly created Partnership Program

  17. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U3Si2-Al and U3Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U3Si2-Al fuel at 4.8 g U/cm3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U3Si-Al with 19.75 % enrichment and U3Si2-Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

  18. Possibilities of Kazakhstan experimental base for space nuclear reactors elements testing

    International Nuclear Information System (INIS)

    To the mid of 70-th in Kazakhstan the surface developing base for space nuclear reactors elements testing was created. The base consists of three test complexes. Two of them - the complexes of test reactors 'Baikal-1' and IGR - are situating on the Semipalatinsk test site, and the third one - complex of WWR-K research reactor - is situating in Alatau village nearby to Almaty city. On 'Baikal-1' and IGR complexes the testings for fuel elements, fuel assemblies, modules and prototypes of nuclear rocket engine reactor and nuclear energetic engine units with turbine-engine energy transmission on the base solid-phase reactor were carrying out. On the WWR-K reactor complex the testing of power generating channels of thermal-emission transmission reactors were conducted. In the paper the assessment of up-to-date experimental base status and it possibilities for further using in space nuclear energy field are given

  19. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  20. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  1. The influence of the furnace design on emissions from small wood pellet burners

    International Nuclear Information System (INIS)

    Two pellet burners have been installed and tested in a small scale boiler for house heating. The boiler is representative for the Swedish households and the burners, upwards and forward burning, are commercially available on the Swedish market. This work focuses on the boiler operation and particularly the potential of improved emissions by changing the furnace design. An insulation of the fireplace lowered the emission of CO by 50% and the emission of OGC by 60% for the upwards burning burner at low load. Modifying the furnace using baffles did not have any influence on the emissions. It is concluded that an increased temperature in the furnace is more important than an increased residence time of the combustible gases to decrease the emissions. At full load both burners emit approximately 300 mg CO per nm3 gas and the emission of OGC are negligible. At half load the emissions of CO increased to 1000 mg/mn3 and OGC to 125 mg/mn3 in the upward burning burner. The forwards burning burner had a small increase in OGC to about 10 mg/mn3 at half load while the emission of CO increased to 800 mg/mn3. The forward burning burner is less influenced on the furnace design compared to the upward burning burner. The comparatively high emissions of OGC for the upward burning burner is explained by the intermittent operation. However, it was possible to reduce the emissions from this burner by ceramic insulation of the furnace Project report from the program: Small scale combustion of biofuels. 3 refs, 12 figs, 2 tab, 1 appendix with 33 figs and 12 tabs

  2. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  3. Integrated infrastructure initiatives for material testing reactor innovations

    International Nuclear Information System (INIS)

    Highlights: → The EU FP7 MTR+I3 project has initiated a durable cooperation between MTR operators. → Improvements in irradiation test device technology and instrumentation were achieved. → Professional training efforts were streamlined and best practices were exchanged. → A framework has been set up to coordinate and optimize the use of MTRs in the EU. - Abstract: The key goal of the European FP6 project MTR+I3 was to build a durable cooperation between Material Testing Reactor (MTR) operators and relevant laboratories that can maintain European leadership with updated capabilities and competences regarding reactor performances and irradiation technology. The MTR+I3 consortium was composed of 18 partners with a high level of expertise in irradiation-related services for all types of nuclear plants. This project covered activities that foster integration of the MTR community involved in designing, fabricating and operating irradiation devices through information exchange, know-how cross-fertilization, exchanges of interdisciplinary personnel, structuring of key-technology suppliers and professional training. The network produced best practice guidelines for selected irradiation activities. This project allowed to launch or to improve technical studies in various domains dealing with irradiation test device technology, experimental loop designs and instrumentation. Major results are illustrated in this paper. These concern in particular: on-line fuel power determination, neutron screen optimization, simulation of transmutation process, power transient systems, water chemistry and stress corrosion cracking, fission gas measurement, irradiation behaviour of electronic modules, mechanical loading under irradiation, high temperature gas loop technology, heavy liquid metal loop development and safety test instrumentation. One of the major benefits of this project is that, starting from a situation of fragmented resources in a strongly competitive sector, it has

  4. Modeling the critical hydrogen concentration in the AECL test reactor

    International Nuclear Information System (INIS)

    Hydrogen is added to a pressurized water reactor (PWR) to suppress radiolysis and maintain reducing conditions. The minimum hydrogen concentration needed to prevent radiolysis is referred to as the critical hydrogen concentration (CHC). The CHC was measured experimentally in the mid-1990s by Elliot and Stuart in a reactor loop at Atomic Energy of Canada (AECL), and was found to be approximately 0.5 scc/kg for typical PWR conditions. This value is well below industry-normal PWR operating levels near 40 scc/kg. Radiation chemistry models have also predicted a low CHC, even below the AECL experimental result. In the last few years some of the radiation chemical kinetic rate constants have been re-measured and G-values have been reassessed by Elliot and Bartels. These new data have been used in this work to revise the models and compare them with AECL experimental data. It is quite clear that the scavenging yields tabulated for high-LET radiolysis by Elliot and Bartels are not appropriate to use in the present context, where track-escape yields are needed to describe the homogeneous recombination kinetics in the mixed radiation field. In the absence of such data for high temperature PWR conditions, we have used the neutron G-values as fitting parameters. Even with this expedient, the model predicts at least a factor of two smaller CHC than was observed. We demonstrate that to recover the reported CHC result, the chemistry of ammonia impurity must be included. - Highlights: ► Hydrogen is added to nuclear reactor cooling loops to prevent radiolysis. ► Tests at AECL were carried out to determine the critical hydrogen concentration. ► Neutron radiolysis G-values need to be modified to understand the results. ► Ammonia impurity needs to be included for quantitative modeling.

  5. Reactor safety research program at Thai test facility

    International Nuclear Information System (INIS)

    Thermal-hydraulics, Hydrogen, Aerosol and Iodine (Thai) aims at providing experimental database for the verification and validation of Lumped Parameter (Lp) and Computational Fluid Dynamics (CFD) codes with 3-dimensional capabilities. Since its construction in 2000, Thai facility has been engaged in the field of reactor safety in the frame of various national (Thai I: 2000-2003, Thai II: 2003-2006, Thai III: 2006-2009, Thai IV: 2009-2012) and international programs (OECD-Thai: 2007-2009). Additionally, experimental data has been provided for several international standard problems (ISP 41, 46, 47 and 49) code validation exercises. Experiments performed in Thai facility cover a wide spectrum or reactor safety relevant issues by investigating separate and coupled-phenomenon experiments under design basis accident and severe-accident-typical scenarios. Experiments are performed in close co-operation with AREVA Erlangen and Grs Koln. Experimental configuration and the operating conditions in Thai vessel typical of those for PWR, BWR and High Temperature Gas Cooled Reactor can be produced thanks to its modular structure, appropriate feeding/generation devices for gases (H2, He, Steam, N2, etc.), Aerosol (inert and hygroscopic), Iodine Radiotracer, and advanced instrumentation. Experiments also cover investigation of passive safety systems, e.g. commercial Par for H2 mitigation in phenomenon orientated experiments to enhance the confidence in the performance of passive mitigation systems during severe accident scenarios and also to establish a common database accessible by a large research community to support further development and validation of the Lp and CFD codes with 3-dimensional capabilities. This paper summarizes experimental investigations made in Thai test facility to investigate issues related to the thermal-hydraulics, fission product (aerosol, iodine) transport and their interaction with containment walls (deposition, resuspension) and passive safety

  6. Core Function Changes from a Breakeven Core to a TRU Burner Core

    International Nuclear Information System (INIS)

    A 600MWe sodium cooled fast reactor named as KALIMER-600 has been developed with a single enrichment fuel. This reactor is a pool-type reactor with a 1,523MW thermal power. The core is loaded with a ternary metallic fuel of 15 w/o TRU enriched TRU-U- 10Zr and it is designed to have breakeven breeding characteristics (CR∼1.0). However, a new demand is how to solve a spent fuel disposal problem because nuclear spent fuel storages shall become full by 2016 year. Therefore, a TRU burner concept which can burn out spent fuel actively is needed instead of a breakeven reactor concept. After all spent fuels from LWRs are burned, another issue may be that a TRU burner can not be operated in a breakeven mode any more. In order to overcome this problem, a new concept, a core function change is proposed in this paper. A reactor will operate as a TRU burner at first and then, will play the role of a breakeven core without any core layout change which does not need TRU supply. Since the nuclear conceptual design of a breakeven core - KALIMER-600 is already finished, TRU burner concepts are based on the KALIMER-600 breakeven core and its safety parameters are asked to be compatible with those of the KALIMER- 600 breakeven core

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  8. Regenerative burner generates more savings

    Energy Technology Data Exchange (ETDEWEB)

    Swinden, D.

    The latest developments in high-efficiency gas-fired burners are traced, and the transfer of the new technology from laboratory to industry is outlined. The system described depends on the ceramic regenerator reducing the flue gas temperature so that conventional cold air fans can be used and on a packing of alumina balls to recover 90% of the available heat in waste gases.

  9. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-11-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility.

  10. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

  11. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  12. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope

    International Nuclear Information System (INIS)

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

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

    International Nuclear Information System (INIS)

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

  14. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  15. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [and others

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  16. Similarity Analysis for Reactor Flow Distribution Test and Its Validation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Ha, Jung Hui [Heungdeok IT Valley, Yongin (Korea, Republic of); Lee, Taehoo; Han, Ji Woong [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The newly derived dimensionless groups are slightly different from Hetsroni's. Reynolds number, relative wall roughness, and Euler don't appear, instead, friction factor appears newly. In order to conserve friction factor Reynolds number and relative wall roughness should be conserved. Since the effect of Reynolds number in high range is small, and since the scaled model is far smaller than prototype the conservation of friction factor is easily obtained by making the model wall just smooth. It is much easier to implement the test design than Hetsroni's because the Reynolds number and relative wall roughness do not appear explicitly. In case that there is no free surface within the interested domain of the reactor, the gravity is of second importance, and in this case the pressure drops should be compensated for in order to compare them between prototype and model. The gravity head compensated pressure drop is directly same to the measured value by a differential pressure transmitter. In order to conserve the gravity effect Froude number should be conserved. In pool type SFR (Sodium Cooled Fast Reactor) there exists liquid level difference, and if the level difference is desired to be conserved, the Froude number should be conserved. Euler number, which represents pressure terms in momentum equation, should be well conserved according to Hetsroni's approach. It is not a wrong statement, but it should be noted that Euler number is NOT an independent variable BUT a dependent variable according to Hong et al. It means that if all the geometrical similarity and the dimensionless numbers are conserved, Euler number is automatically conserved. So Euler number need not be considered in case that the perfect geometrical similarity is kept. However, even in case that the geometrical similarity is not conserved, it possible to conserved the velocity field similarity by just conserve Euler number. It gives tolerance to the engineer who designs the test

  17. Similarity Analysis for Reactor Flow Distribution Test and Its Validation

    International Nuclear Information System (INIS)

    The newly derived dimensionless groups are slightly different from Hetsroni's. Reynolds number, relative wall roughness, and Euler don't appear, instead, friction factor appears newly. In order to conserve friction factor Reynolds number and relative wall roughness should be conserved. Since the effect of Reynolds number in high range is small, and since the scaled model is far smaller than prototype the conservation of friction factor is easily obtained by making the model wall just smooth. It is much easier to implement the test design than Hetsroni's because the Reynolds number and relative wall roughness do not appear explicitly. In case that there is no free surface within the interested domain of the reactor, the gravity is of second importance, and in this case the pressure drops should be compensated for in order to compare them between prototype and model. The gravity head compensated pressure drop is directly same to the measured value by a differential pressure transmitter. In order to conserve the gravity effect Froude number should be conserved. In pool type SFR (Sodium Cooled Fast Reactor) there exists liquid level difference, and if the level difference is desired to be conserved, the Froude number should be conserved. Euler number, which represents pressure terms in momentum equation, should be well conserved according to Hetsroni's approach. It is not a wrong statement, but it should be noted that Euler number is NOT an independent variable BUT a dependent variable according to Hong et al. It means that if all the geometrical similarity and the dimensionless numbers are conserved, Euler number is automatically conserved. So Euler number need not be considered in case that the perfect geometrical similarity is kept. However, even in case that the geometrical similarity is not conserved, it possible to conserved the velocity field similarity by just conserve Euler number. It gives tolerance to the engineer who designs the test

  18. Establishing a safety and licensing basis for generation IV advanced reactors. License by test

    International Nuclear Information System (INIS)

    The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

  19. Reactor pressure vessel design of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The reactor pressure vessel (RPV) of the HTTR is 5.5m (inside diameter), 13.2m (inside height), and 122mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr-1Mo steel is chosen as the material for RPV. The temperature reaches about 400 deg. C at normal operation. The fluence of the RPV is estimated to be less than 1 x 1017n/cm2 (E > 1MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr-1Mo steel for the RPV; embrittlement parameters, J-factor and X-bar are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr-1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and X-bar , and nil-ductility transition temperatures, TNDT, by drop weight tests, are shown. In-service inspection and results of R and Ds are also described

  20. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  1. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  2. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  3. Disposition of weapon-grade plutonium with pebble bed type HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. Furthermore, a method was found that the power coefficient could be made negative by heavy Pu loading in the Pu burner ball fuels

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

    International Nuclear Information System (INIS)

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

  5. Development and verification test of integral reactor major components

    International Nuclear Information System (INIS)

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

  6. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  7. Communication of Advanced Test Reactor probabilistic risk assessment results

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 results were efficiently communicated in two reports following the completion of revision 1 of the ATR PRA. As the ATR PRA including external events fills four large volume, it was considered impractical to expect all of the individuals at ATR who could benefit from the information to read the entire PRA. Even though many ATR personnel received training in PRA methodology and were involved in some aspects of the PRA, another hinderance to effective communication of the PRA results is that the PRA was written and organized to meet the requirements of practitioners and reviewers who are well-versed in PRA methods. Therefore, two PRA summary reports, an ATR risk summary report and an ATR functional group summary report, were written to communicate the ATR PRA results and insights to interested ATR personnel

  8. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  9. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  10. Qualification in the reactor Siloe of low enriched fuels for research and test reactors

    International Nuclear Information System (INIS)

    For nearly two years now, in the scope of the Reduced Enrichment Research and Test Reactor (RERTR) Program (CEA-ANL-CERCA Agreements), low enriched fuel has been irradiated in Siloe. In 1981 a complete 45% enriched fuel element (U Alx compound) was irradiated. A burn-up of 50% was obtained without any difficulty. Since June 1982 4 U3Si fuel plates are being irradiated. These plates, with a density of 5.5g of total uranium per cubic centimeter (two plates), and 6.0g per cubic centimeter (the other two plates) have already reached a mean burn-up of about 20% and their behaviour up till now is excellent. The fuel element and the plates have been manufactured by CERCA

  11. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors)

  12. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  13. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 2000C. The design description and results of the prototype capsule performance are presented

  14. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  15. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

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

  17. 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.

    1998-02-01

    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.

  18. Removal of the Plutonium Recycle Test Reactor - 13031

    International Nuclear Information System (INIS)

    The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associated underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the

  19. Personnel Monitoring Practices at the National Reactor Testing Station

    International Nuclear Information System (INIS)

    The film badge utilized in the personnel monitoring programme at the National Reactor Testing Station (NRTS) enables one to determine the type of radiation, estimate the average energy of the gamma component and evaluate the dose to the badge in the range of ∼30 mrad to 800 rad. This range is adequate for most accidental as well as routine exposure monitoring. In emergency situations, an estimated beta-gamma exposure dose can be provided in less than 30 minutes following receipt of the film by the Dosimetry Branch. A personnel neutron threshold detector (PNTD) in each badge provides the capability of rapid field screening of personnel following a criticality incident. The PNTD offers a sensitive and reliable method for the detection and estimation of thermal ( 2.5 MeV) neutrons under a variety of conditions. Modifications currently under way should also provide for estimation of epithermal (0.4- 1.4 eV) neutrons and provide a detection limit at time-zero of <0.2 mrad thermal or epithermal. Fixed nuclear accident dosimeters (NAD) are strategically located throughout the various operating areas. These contain elemental foils for determination of neutron flux and estimation of the spectrum and chemical dosimeters for measuring the gamma flux. Normalization of the PNTD to the NAD can be accomplished to assist in determining orientation of personnel and calculation of dose. The data obtained during the recent SPERT I and SNAPTRAN-3A reactor destruct series indicated good correlation between the PNTD and similar data obtained from the NAD system. A computer and associated peripheral punched-card equipment are utilized at the NRTS for the dose determinations, maintenance of accurate health-physics records and the rapid retrieval of radiation exposure information in emergency situations such as the SL-1 reactor accident. During a six-week period complete updated exposure lists were furnished the control group each morning at 8 a.m. covering all exposures received at SL-1

  20. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    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.

  2. Testing of HTR UO2 TRISO fuels in AVR and in material test reactors

    International Nuclear Information System (INIS)

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO2 TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO2 TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO2 TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C

  3. OPTIMIZATION OF COAL PARTICLE FLOW PATTERNS IN LOW NOX BURNERS

    Energy Technology Data Exchange (ETDEWEB)

    Jost O.L. Wendt; Gregory E. Ogden; Jennifer Sinclair; Stephanus Budilarto

    2001-08-20

    The proposed research is directed at evaluating the effect of flame aerodynamics on NO{sub x} emissions from coal fired burners in a systematic manner. This fundamental research includes both experimental and modeling efforts being performed at the University of Arizona in collaboration with Purdue University. The objective of this effort is to develop rational design tools for optimizing low NO{sub x} burners to the kinetic emissions limit (below 0.2 lb./MMBTU). Experimental studies include both cold and hot flow evaluations of the following parameters: flame holder geometry, secondary air swirl, primary and secondary inlet air velocity, coal concentration in the primary air and coal particle size distribution. Hot flow experiments will also evaluate the effect of wall temperature on burner performance. Cold flow studies will be conducted with surrogate particles as well as pulverized coal. The cold flow furnace will be similar in size and geometry to the hot-flow furnace but will be designed to use a laser Doppler velocimeter/phase Doppler particle size analyzer. The results of these studies will be used to predict particle trajectories in the hot-flow furnace as well as to estimate the effect of flame holder geometry on furnace flow field. The hot-flow experiments will be conducted in a novel near-flame down-flow pulverized coal furnace. The furnace will be equipped with externally heated walls. Both reactors will be sized to minimize wall effects on particle flow fields. The cold-flow results will be compared with Fluent computation fluid dynamics model predictions and correlated with the hot-flow results with the overall goal of providing insight for novel low NO{sub x} burner geometry's.

  4. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  5. Large break LOCA experiment at reactor thermohydraulic test loop

    International Nuclear Information System (INIS)

    The experiments of large break LOCA in the reactor Thermohydraulic Test Loop (UUTR) has been done. The experiments were held at hot leg side by use of accumulator safety injection system and without accumulator. Two experiments were done without activating the high and low pressure safety injection system, while make up system was activated until the experiments were stopped. The test were done at 1 MWt power with about 9,4 kg/sec primary coolant flow rate, and pressure 154 bar. The phenomena were only be limited on the effect of accumulator to the system during LOCA. The results of the experiments indicated a similar system depressurization phenomena for both with and without accumulator was activated. The system trip were happened at very closely different time for both at around 10 t h cycle. The significant difference were that in the experiments were the accumulator was activated, the system depressurization was going more slowly than without the accumulator, and the rods and the fluid temperature were more lower. In the case, the water injection system from the accumulator was able to reduced the rod and the cooling temperature as long as the water inventory is available

  6. Design of the TFTR [Tokamak Fusion Test Reactor] maintenance manipulator

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) plans to generate a total of 3 x 1021 neutrons during its deuterium-tritium run period in 1900. This will result in high levels of radiation, especially within the TFTR vacuum vessel. The maintenance manipulator's mission is to assist TFTR in meeting Princeton Plasma Physics Laboratory's personnel radiation exposure criteria and in maintaining as-low-as-reasonably-achievable principals by limiting the radiation exposure received by operating and maintenance personnel. The manipulator, which is currently being fabricated and tested by Kernforschungszentrum Karlsruhe, is designed to perform limited, but routine and necessary, functions within the TFTR vacuum torus after activation levels within the torus preclude such functions being performed by personnel. These functions include visual inspection, tile replacement, housekeeping tasks, diagnostic calibrations, and leak detection. To meet its functional objectives, the TFTR maintenance manipulator is required to be operable in TFTR's very high vacuum environment (typically 2 x 10-8 Torr). It must also be bakeable at 150 degree C and able to withstand the radiation environment

  7. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item`s test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship`s demand. (author).

  8. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment); Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu.

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author).

  9. Computational fluid dynamics in oil burner design

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

    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.

  10. A numerical investigation of the aerodynamics of a furnace with a movable block burner

    OpenAIRE

    T.J. Fudihara; L. Goldstein Jr.; Mori, M.

    2007-01-01

    In this work the air flow in a furnace was computationally investigated. The furnace, for which experimental test data are available, is composed of a movable block burner connected to a cylindrical combustion chamber by a conical quarl. The apertures between the movable and the fixed blocks of the burner determine the ratio of the tangential to the radial air streams supplied to the furnace. Three different positions of the movable blocks were studied at this time. A three-dimensional invest...

  11. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  12. Models for transient analyses in advanced test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gabrielli, Fabrizio

    2011-02-22

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are

  13. Gasifier burner ignition system

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    The BI-GAS pilot plant is designed such that its lower region (Stage I) is the combustion zone where oxygen and steam contact and react with recycled char. As with other combustion systems, provisions must be made to initiate combustion at start-up, to reinitiate combustion in case of a process upset where combustion is lost, and to continuously monitor the presence of combustion. An ignition system had to be developed, capable of reliable and repeated operation at pressures up to 1500 psi in a methane-rich or otherwise reducing atmosphere. The initial development work was done by Babcock and Wilcox and included development of both the ignitor system and the flame confirmation system. B and W's initial proposal specifically dealt with investigating a hypergolic (chemical auto-combustion) igniter. Hypergolic ignition is the spontaneous combustion of a compound upon contact with an oxygen containing media. This oxygen source includes air, oxygen, carbon dioxide, and water. The liquid compound studied was triethylaluminum (Al(C/sub 2/H/sub 5/)/sub 3/) otherwise identified as TEA and supplied by Ethyl Corporation, Baton Rouge, Louisiana. The hypergolic ignition system has been operated successfully and proved reliable at high pressure (750 psig) through repeated testing over a three-year period. The system designed by Stearns-Roger based on the study by Babcock and Wilcox was basically correct. Two relatively minor design defects and operational revisions to improve performance were accomplished by on-site personnel with little expenditure of time or money. The remaining problems currently experienced with the TEA ignition system are considered minor. Further work should continue to determine the lowest possible TEA concentration that can be used and still provide consistent ignition, and the system should be tested soon at the full design operating pressure of 1500 psig.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

  16. RERTR-2004: International meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Book of abstracts

    International Nuclear Information System (INIS)

    Oral and poster presentations of the Meeting covered the following topics: National and international programs related to Reduced Enrichment for Research and Test Reactors (RERTR); development of new fuel types, testing, fabrication, modelling; studies of reactor cores conversion from highly enriched to low enriched fuel, including licensing; new and converted reactors; spent fuel management including storage and transportation; production of Molybdenum 99 under converted core conditions

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

    International Nuclear Information System (INIS)

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

  18. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM

  19. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  20. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  1. Hydraulic stand for testing the models of nuclear reactors

    International Nuclear Information System (INIS)

    Paper describes the basic design and hydraulic characteristics of the MR experimental bench including equipment and coolant circulation circuit, two-loop reactor hydraulic model and measuring system. The bench enables to investigate into thermal and hydraulic characteristics of models of two- and four-loop vessel reactors. Paper dwells upon research and training use of the bench

  2. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  3. High field, low current operation of engineering test reactors

    International Nuclear Information System (INIS)

    Steady state engineering test reactors with high field, low current operation are investigated and compared to high current, lower field concepts. Illustrative high field ETR parameters are R = 3 m, α ∼ 0.5 m, B ∼ 10 T, β = 2.2% and I = 4 MA. For similar wall loading the fusion power of an illustrative high field, low current concept could be about 50% that of a lower field device like TIBER II. This reduction could lead to a 50% decrease in tritium consumption, resulting in a substantial decrease in operating cost. Furthermore, high field operation could lead to substantially reduced current drive requirements and cost. A reduction in current drive source power on the order of 40 to 50 MW may be attainable relative to a lower field, high current design like TIBER II implying a possible cost savings on the order of $200 M. If current drive is less efficient than assumed, the savings could be even greater. Through larger β/sub p/ and aspect ratio, greater prospects for bootstrap current operation also exist. Further savings would be obtained from the reduced size of the first wall/blanket/shield system. The effects of high fields on magnet costs are very dependent on technological assumptions. Further improvements in the future may lie with advances in superconducting and structural materials

  4. Combustion Characteristics of Oxy-fuel Burners for CO2 Capturing Boilers

    Science.gov (United States)

    Ahn, Joon; Kim, Hyouck Ju; Choi, Kyu Sung

    Oxy-fuel boilers have been developed to capture CO2 from the exhaust gas. A 50 kW class model burner has been developed and tested in a furnace type boiler. The burner has been scaled up to 0.5 and 3 MW class for fire-tube type boilers. The burners are commonly laid out in a coaxial type to effectively heat the combustion chamber of boilers. Burners are devised to support air and oxy-fuel combustion modes for the retrofitting scenario. FGR (flue gas recirculation) has been tried during the scale-up procedure. Oxy-fuel combustion yields stretched flame to uniformly heat the combustion chamber. It also provides the high CO2 concentration, which is over 90% in dry base. However, pure oxy-fuel combustion increases NO concentration, because of the reduced flow rate. The FGR can suppress the thermal NOx induced by the infiltration of the air.

  5. Heat transfer and combustion characteristics of a burner with a rotary regenerative heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, Yasuo; Kaji, Hitoshi; Arai, Norio

    1998-07-01

    The authors have developed a Rotary Regenerative Combustion (RRX) System, which is coupled with a compact high efficiency regenerative air heat exchanger and a combustion burner. This system contributes to saving energy of fuel firing industrial furnaces and decreases NO{sub x} emission. This technology can be considered as a solution of greenhouse problem. This paper, discusses a compact high efficiency regenerative air heat exchanger in comparison with the existing types of regenerative burners and reverse firing with high momentum fuel jet (with motive fluid) in the furnace. This burner is compact in size, with high fuel efficiency, low NOx emission, easy to operate, and reliable, based on the results of field tests and commercial operations. The authors can say that the RRX system is a regenerative burner of the second generation.

  6. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    International Nuclear Information System (INIS)

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  7. Refurbish research and test reactors corresponding to global age of nuclear energy

    International Nuclear Information System (INIS)

    This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

  8. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  9. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  10. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author)

  11. Reacteur Jules Horowitz (RJH): A new material testing reactor. Status report

    International Nuclear Information System (INIS)

    The 'REACTEUR JULES HOROWITZ' (RJH) is a new research reactor dedicated to material and nuclear fuel testing. This reactor, which will be erected in the CEA Cadarache nuclear research Center is now at a feasibility study stage. At the beginning of the next century, at a time when most of existing material testing reactors will have to be shutdown or will be at the end of their lifetime, le RJH will offer outstanding neutron flux levels (twice those of existing french reactors). This paper deals with the following topics - functional specifications of the project, - safety approach, - design and construction codes, - alternative designs under consideration at the feasibility stage. (author)

  12. Experimental study of porous metal burners for domestic stove applications

    International Nuclear Information System (INIS)

    Highlights: • The flat flame cooktop burner is experimental and analytical investigated. • The heat transfer flux can be evaluated using analytical and numerical methods. • The performance of the flat flame burners is superior to Bunsen flame type burners. • Efficiency and emissions of the flat flame burners can be well controlled. - Abstract: This paper investigates a clean and highly efficient domestic stove burner composed of a flat flame burner for cooking and water heating. The feasibility of the flat flame burner is experimentally verified by demonstrating that the flame is stabilized by a porous metal medium and by comparison with a typical Bunsen flame burner. The flame appearance, temperature distribution, relative thermal efficiency and pollution emissions in terms of Emission Index of CO (EICO) and Emission Index of NOx (EINOx) were measured and analyzed. The results show that the operating range, turndown ratio, and pollution emissions of the flat flame burners are superior to those of traditional Bunsen flame burners. The heat transfer and efficiency for both the jet flame burner and the flat flame burner can be evaluated using analytical and numerical methods. Furthermore, in contrast to traditional Bunsen flame burners, the efficiency and pollution emissions of flat flame burners are not strongly affected by the distance between the cool boundary of pot or pan and the burner exit. For domestic stove applications in particular, where different sized pots and pans are used, the efficiency and pollution emissions can be well controlled with a flat flame burner

  13. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  14. Development and demonstration of a gas-fired recuperative confined radiant burner (deliverable 42/43). Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-06-01

    The objective of the project was to develop and demonstrate an innovative, efficient, low-pollutant, recuperative gas-fired IR-system (infrared radiation) for industrial processes (hereafter referred to as the CONRAD-system). The CONRAD-system is confined, so flue gases from the combustion can be kept separated from the product. The gas/air mixture to the burner is preheated by means of the flue gas, which increases the radiant efficiency of the CONRAD-system significantly over traditional gas-fired IR burners. During the first phase of the project, the CONRAD-system was designed and developed. The conducted work included a survey on suitable burner materials, modelling of the burner system, basic design of burner construction, control etc., experimental characterisation of several preprototypes and detailed design of the internal heat exchanger in the burner. The result is a cost effective burner system with a documented radiant efficiency up to 66% and low emissions (NO{sub x} and CO) all in accordance with the criteria of success set up at the start of the project. In the second phase of the project, the burner system was established and tested in laboratory and in four selected industrial applications: 1) Drying of coatings on sand cores in the automotive industry. 2) Baking of bread/cake. 3) General purpose painting/powder curing process 4. Curing of powder paint on wood components. The results from the preliminary tests Overe used to optimise the CONRAD-system, before it was applied in the industrial processes and demonstrated. However, the optimised burners manufactured for demonstration suffered from different 'infant failures', which made the installation in an industrial environment very cumbersome, and even impossible in the food industry and the automotive industry. In the latter cases realistic laboratory tests Overe carried out and the established know how reported for use when the burner problems are overcome.(au)

  15. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  16. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The United States Department of Energy's Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation

  17. An intelligent monitoring system for the detection of slag deposition on a pulverized coal fired burner

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.K.; Wilcox, S.J.; Ward, J.; Lewitt, M. [University of Glamorgan, Pontypridd (United Kingdom). School for Technology

    2005-08-01

    The objective of this paper is to describe the further development of a monitoring system to detect the presence of so-called burner eyebrows, i.e. relatively large deposits of slag around the burner quarl in pulverized coal fired boilers. Experiments were undertaken with a range of coals and with various artificial eyebrows constructed from cast refractory inserts. The system uses a microphone to detect combustion noise and an infrared sensor which measures flame radiation, and the signals from these cheap, easily installed sensors were analyzed by a hybrid neural network. In tests with two coals, the system was able to distinguish the different eyebrows with a high degree of accuracy if representative data were used to train the network for each particular coal. In further tests with a range of six different coals, the system was able to distinguish between a clean burner and one fitted with a particular sized eyebrow. In this case, it proved to be possible to use only the features from three of the coals in the training process and the data from the remaining fuels for validation. The monitoring system, therefore, appears to be relatively independent of changes to the coal fired by the burner if trained with a representative range of coals. Finally, this paper presents a possible method to detect burner eyebrows via the evaluation of so-called 'eyebrow indices' using a self-organizing map which is trained solely using clean burner sensor patterns.

  18. Operating safety experience of fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Operational safety criteria for nuclear reactors are very stringent and it is essential to incorporate adequate inherent and engineered safety features in the design to ensure safe operation of the reactor. Commissioning and operation of FBTR, being first of its kind in India based on nuclear chain reaction maintained by fast neutrons and use of high temperature liquid sodium as coolant, was a challenging task. Safe operation of the reactor for the past 17 years with good performance of sodium systems and the indigenous plutonium rich carbide fuel, touching a burn up level of 100 GWd/t has underlined the high level of design and operation competence achieved

  19. Testing the AVR for theoretical models of reactor physics

    International Nuclear Information System (INIS)

    More than 20 years' operation of the AVR was used in two ways to check theoretical models of reactor physics: The accompanyment and pre-calculation of normal operating behaviour and extensive experimental programs on special problems have contributed in an excellent way to working out and validating large reactor simulation models for the HTR. The large number of types of fuel elements used and the relatively complex geometry of this reactor make great demands of the theoretical models and have led to the experience that only a combination of the physical phenomena from neutron, heat and fission material transport will lead to a satisfactory description of reality. (orig.)

  20. A walkthrough of the Copenhagen atomics waste burner design

    International Nuclear Information System (INIS)

    This talk will give an introduction on Copenhagen Atomics and then elaborate on Copenhagen Atomics major objectives, which is to build thorium molten salt reactors (MSR) on an assembly line preferably with the reactor core fitted in a 40 foot shipping container. The first model will be 50 MWt and it will start on a 78% LiF-22% ThF4 salt, mixed with plutonium and actinides from spent nuclear fuel (SNF). Over decades 233U in the salt will increase and eventually help to burn out long lived actinides. The plan is to locate these waste burners at the site of existing Light Water Reactors (LWR) plants to avoid SNF transport, avoid approval of new sites and take advantage of the security and power grid infrastructures. The talk will allow the audience to understand similarities and differences between the Copenhagen Atomics Waste Burner and other MSRs such as LFTR and IMSR. The main objective of Copenhagen Atomics is to convince the public that it is possible to build a machine (MSR), which can burn many of the long lived actinides out of SNF and reduce the storage time from 100.000+ years to 300 years, while at the same time produce enough energy to pay for the process and decommissioning. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste. (author)

  1. 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)

    1996-12-31

    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.)

  2. Development of a multi-fuel burner for coal gasification process; Entwicklung eines Kombibrenners fuer den Kohlevergasungsprozess

    Energy Technology Data Exchange (ETDEWEB)

    Al-Halbouni, Ahmad; Rahms, Hendrik; Chalh-Andreas, Bachir [Brinkmann Industrielle Feuerungssysteme GmbH, Voerde (Germany); Giese, Anne [Gas- und Waerme-Institut Essen e.V., Essen (Germany); Benim, Ali Cemal [Fachhochschule Duesseldorf (Germany)

    2013-08-15

    In the course of a German ZIM cooperative research project, Brinkmann Industrielle Feuerungssysteme GmbH develops a supersonic oxygen-multi-fuel burner in close cooperation with its research partners Gas- und Waerme-Institut essen e.V. (GWI) and Duesseldorf University of Applied Sciences (FHD). This burner is capable of combusting natural gas as well as light oil efficiently, using pure oxygen as an oxidizer. It is intended to be used primarily for energy-intensive applications, but especially as a start-up burner for coal gasification processes. In these processes, specific operating conditions can be found, such as fluctuating pressures, high temperatures and inert atmospheres. Therefore, the main goal of the development is aimed at utilizing the high energy densities found in supersonic by oxy-fuel combustion. This article covers several burner development phases, from initial design and manufacturing activities to burner testing and optimisation. Results achieved up to now are presented and next steps defined. (orig.)

  3. Experimental capabilities of the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service

  4. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  5. Nondestructive testing of ampoules with lithium ceramics designed for blanket of thermonuclear reactor

    International Nuclear Information System (INIS)

    Full text: There are carried out prolonged radiation tests on research reactor WWR-K of ceramic materials made of lithium titanate Li2Ti03 with enrichment 36Li to 90 % manufactured in the form of sintered small balls and cylinder tablets put in experimental assembles (ampoules). At the present time tritium titanate is considered as one of the possible candidates of tritium production zone for demonstration international thermonuclear reactor blanket. Before feeding into the reactor experimental assemblages with Li2TiO3 were exposed to nondestructive control on horizontal channel of reactor with 'Agava' plant use by the neutron radiography method. The purpose of this work is on the one hand feeding quality control of tablets and small balls of lithium ceramics into experimental assembles, on the other hand the efficiency test of neutron radiography plant work after long stoppage of WWR-K reactor and the geometry change of irradiative channels and active zone of reactor

  6. Testing mass-varying neutrinos with reactor experiments

    OpenAIRE

    Schwetz, Thomas; Winter, Walter

    2005-01-01

    We propose that reactor experiments could be used to constrain the environment dependence of neutrino mass and mixing parameters, which could be induced due to an acceleron coupling to matter fields. There are several short-baseline reactor experiment projects with different fractions of air and earth matter along the neutrino path. Moreover, the short baselines, in principle, allow the physical change of the material between source and detector. Hence, such experiments offer the possibility ...

  7. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  8. Operation and maintenance experience with control rod and their drive mechanisms of fast breeder test reactor

    International Nuclear Information System (INIS)

    This paper explains the functional and construction features of Control Rod Drive Mechanism (CRDM) and control rod used in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type sodium cooled fast reactor. It discusses all safety related incidents and failures encountered during its service in reactor, the solutions evolved and modifications carried out to prevent recurrence. It also details the maintenance activities and periodical surveillance carried out. The results of a reliability analysis done are also discussed. (author)

  9. Regenerative burner combination and method of burning a fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wills, B.J.

    1992-06-17

    Regenerative burners fire alternatively into respective radiant tubes which are closed at their ends remote from the burners. Products of combustion from each flame tube pass to the closed end of the tube and back to be exhausted from the radiant tube associated with the firing burner through a transfer duct to the other burner, where heat is extracted before the products of combustion are discharged, for heating combustion air for use when the other burner is firing. (Author).

  10. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  11. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  12. Rotrix `vortex breakdown` burner turbulence-stabilized combustion of heating oil

    Energy Technology Data Exchange (ETDEWEB)

    Hofbauer, P. [Viessmann Manufacturing Co., Inc., Ontario (Canada)

    1995-04-01

    For the past two years, the Viessmann MatriX radiant burner has been setting the standard for low emission combustion of gas. Now, with the RotriX burner, Viessmann has succeeded in drastically reducing nitrogenoxide emissions in the combustoin of oil. After a successful test period, the RotriX burner is now being introduced to the market. The RotriX oil burner consequently takes into account the mechanisms in the creation of harmful emissions in the combustion of heating oil No. 2, and guarantees stable combustion under any operating conditions. The burner has the following features: heating oil is combusted only after complete vaporization and mixing with combustion air and recirculated flue gases; the flame is not stabilized with a turbulator disk, but a strong turbulating current is created by means of the Vortex Breakdown phenomenon, which develops a very stable flame under any operating conditions; and high internal flue gas recirculation rates lower the flame temperature to the point where thermal NO formation is reduced to the same low level as in the combustion of gas. The new burner has extremely low emissions of NOx < 60 mg/kWh, and CO < 5 mg/kWh at a CO{sub 2} concentraiton of 14%.

  13. Design Strategy and Constraints for Medium-Power Lead-Alloy-Cooled Actinide Burners

    International Nuclear Information System (INIS)

    We outline the strategy and constraints adopted for the design of medium-power lead-alloy-cooled actinide-burning reactors that strive for a lower cost than accelerator-driven systems and for robust safety. Reduced cost is pursued through the use of (1) a modular design and maximum power rating to capitalize on an economy of scale within the constraints imposed by modularity, (2) a very compact and simple supercritical-CO2 power cycle, and (3) simplifications of the primary system allowed by the use of lead coolant. Excellent safety is pursued by adopting the integral fast reactor approach of achieving a self-controllable reactor that responds to all key abnormal occurrences, including anticipated transients without scrams, by a safe shutdown without exceeding core integrity limits. The three concepts developed are the fertile-free actinide burner for incineration of all transuranics from light water reactor (LWR) spent fuel, the fertile-free minor actinide (MA) burner for preferential burning of MAs working in tandem with LWRs or gas-cooled thermal reactors, and the actinide burner with thorium fuel aimed also at reducing the electricity generation costs through longer-cycle operation

  14. Deep-Burner DB-MHR: physics and computation

    International Nuclear Information System (INIS)

    The paper summarizes the studies on the Deep-Burner Modular Helium Reactor (DBMHR) concept-design of General Atomics, which have been carried-out by FRAMATOME-ANP, General Atomics and Entergy, with the valuable support of CEA, in the framework of a joint collaboration on the Reactor-Based Transmutation Program. Preliminary design studies as well as sensitivity studies and fuel-cycle studies performed both with probabilistic and deterministic methodology are described. Emphasis is put on most attractive physical and computational aspects. A survey on the current investigation on the design uncertainties, the future search for ways to improve the transmutation worth in a double-stratum strategy, and the computational tools improvement are also presented. (authors)

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

    International Nuclear Information System (INIS)

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

  16. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  17. Sipping Test: Checking for Failure of Fuel Elements at the OPAL Reactor

    International Nuclear Information System (INIS)

    Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented

  18. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of Keff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  19. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of Keff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  20. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    International Nuclear Information System (INIS)

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs

  1. Neutronic calculation for cobalt irradiation devices and test loop of etrr2 research reactor

    International Nuclear Information System (INIS)

    MCNP monte Carlo code were used to model ETRR-2 research reactor. The model were used to simulate the irradiation facilities of the reactor. The reactivity worth and neutron flux for cobalt irradiation device and test loop were calculated. The axial thermal flux were also calculated at the six irradiation water channels around the reactor core. The results of the present model were compared both with the experimental measurements for cobalt device and with the design calculations for test loop. Satisfactory agreement were found between the present MCNP results and measurements for both cobalt device and test loop

  2. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  3. Low NO[sub x] regenerative burner

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1992-12-01

    A joint development project between British Gas and Hotwork Development has resulted in maintaining the efficiency of a regenerative burner but without the penalty of the higher NO[sub x] emissions normally associated with combustion air preheat. (author)

  4. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO2) mixed with urania (UO2). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified

  5. Regenerative burner use on reheat furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Baggley, G.W. [Bloom Engineering Co. Inc., Pittsburgh, PA (United States)

    1995-06-01

    The environmental advantages of using regenerative burner technology on steel reheat furnaces are explored in this article, in particular improved fuel energy efficiencies and reduced pollution emissions, of nitrogen oxides and carbon monoxide. Experience of the use of regenerative burners in the United States and Japan, where they have achieved significant market penetration is also described, including a case history of a top-fired billet reheat furnace installed in the United States. (UK)

  6. Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

    International Nuclear Information System (INIS)

    The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

  7. Outline of reactor physics tests conducted during the power-up tests in the nuclear ship MUTSU

    Energy Technology Data Exchange (ETDEWEB)

    Itagaki, Masafumi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Gakuhari, Kazuhiko; Okada, Noboru

    1992-11-01

    The present report describes the outline of a series of reactor physics tests conducted during the power-up tests in the nuclear ship MUTSU, which started on March 29, 1990. The basic physics design parameters have been confirmed from these test results. In spite of a 16 year reactor stop and refabrications of fuel assemblies and control rods in 1989, no change in reactor physics performance through the long period has been also demonstrated by comparing the present measured results with the data received in 1974. The digital reactivity meter used in the physics tests enabled us to perform more efficient and accurate measurements of reactivity than the conventional period method. Most of the physics test results show the three-dimensional (3-D) core characteristics peculiar to ship reactors which are caused by partial insertion of control rods even under full power conditions. A 3-D reactor physics code developed for the MUTSU reactor has given excellent calculation results which agree quite well with these measured characteristics. (author).

  8. Outline of reactor physics tests conducted during the power-up tests in the nuclear ship MUTSU

    International Nuclear Information System (INIS)

    The present report describes the outline of a series of reactor physics tests conducted during the power-up tests in the nuclear ship MUTSU, which started on March 29, 1990. The basic physics design parameters have been confirmed from these test results. In spite of a 16 year reactor stop and refabrications of fuel assemblies and control rods in 1989, no change in reactor physics performance through the long period has been also demonstrated by comparing the present measured results with the data received in 1974. The digital reactivity meter used in the physics tests enabled us to perform more efficient and accurate measurements of reactivity than the conventional period method. Most of the physics test results show the three-dimensional (3-D) core characteristics peculiar to ship reactors which are caused by partial insertion of control rods even under full power conditions. A 3-D reactor physics code developed for the MUTSU reactor has given excellent calculation results which agree quite well with these measured characteristics. (author)

  9. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

  10. Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

    International Nuclear Information System (INIS)

    A preliminary design of a uranyl nitrate test reactor is developed, with emphasis placed on the core neutronics and cross section development. ENDF/B-IV cross section data and the AMPX system were used to develop a 25 group neutron cross section library. A series of one-dimensional transport calculations were made in order to arrive at a reference design. Power densities of 16.5 Kw/1 appear to be attainable in the 217 pin FFTF test subassembly, with a peak neutron flux in the test zone of 2.4 x 1014 n/cm2-sec. Other engineering features pertinent to the overall system design are discussed, including: (1) corrosion, (2) treatment of radiolytic gas, (3) heat removal, and (4) reactor control

  11. Development of research reactor simulator and its application to dynamic test-bed

    International Nuclear Information System (INIS)

    We developed a real-time simulator for 'High-flux Advanced Neutron Application ReactOr (HANARO), and the Jordan Research and Training Reactor (JRTR). The main purpose of this simulator is operator training, but we modified this simulator into a dynamic test-bed (DTB) to test the functions and dynamic control performance of reactor regulating system (RRS) in HANARO or JRTR before installation. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The software includes a mathematical model that implements plant dynamics in real-time, an instructor station module that manages user instructions, and a human machine interface module. The developed research reactor simulators are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by actual RRS cabinet, and was interfaced using a hard-wired and network-based interface. RRS cabinet generates control signals for reactor power control based on the various feedback signals from DTB, and the DTB runs plant dynamics based on the RRS control signals. Thus the Hardware-In-the-Loop Simulation between RRS and the emulated plant (DTB) has been implemented and tested in this configuration. The test result shows that the developed DTB and actual RRS cabinet works together simultaneously resulting in quite good dynamic control performances. (author)

  12. Reduced enrichment for research and test reactors. Proceedings

    International Nuclear Information System (INIS)

    The 12th meeting was attended by 113 participants coming from 21 countries and from EURATOM and IAEA.42 reports were presented orally within 10 sessions dealing with 5 main topics: 1) programs(5); 2) fuels(12); 3) reactor conversions(17); 5) high performance neutron sources(4); 5) others(4). (HP)

  13. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included

  14. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  15. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  16. Combustion Characteristics of Butane Porous Burner for Thermoelectric Power Generation

    Directory of Open Access Journals (Sweden)

    K. F. Mustafa

    2015-01-01

    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.

  17. Study and mathematical model of ultra-low gas burner

    International Nuclear Information System (INIS)

    The main objective of this project is prediction and reduction of NOx and CO2 emissions under levels recommended from European standards for gas combustion processes. A mathematical model of burner and combustion chamber is developed based on interacting fluid dynamics processes: turbulent flow, gas phase chemical reactions, heat and radiation transfer The NOx prediction model for prompt and thermal NOx is developed. The validation of CFD (Computer fluid-dynamics) simulations corresponds to 5 MWI burner type - TEA, installed on CASPER boiler. This burner is three-stream air distribution burner with swirl effect, designed by ENEL to meet future NOx emission standards. For performing combustion computer modelling, FLUENT CFD code is preferred, because of its capabilities to provide accurately description of large number of rapid interacting processes: turbulent flow, phase chemical reactions and heat transfer and for its possibilities to present wide range of calculation and graphical output reporting data The computational tool used in this study is FLUENT version 5.4.1, installed on fs 8200 UNIX systems The work includes: study the effectiveness of low-NOx concepts and understand the impact of combustion and swirl air distribution and flue gas recirculation on peak flame temperatures, flame structure and fuel/air mixing. A finite rate combustion model: Eddy-Dissipation (Magnussen-Hjertager) Chemical Model for 1, 2 step Chemical reactions of bi-dimensional (2D) grid is developed along with NOx and CO2 predictions. The experimental part of the project consists of participation at combustion tests on experimental facilities located in Livorno. The results of the experiments are used, to obtain better vision for combustion process on small-scaled design and to collect the necessary input data for further Fluent simulations

  18. Enhanced Combustion Low NOx Pulverized Coal Burner

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

    economic evaluation and commercial application. During the project performance period, Alstom performed computational fluid dynamics (CFD) modeling and large pilot scale combustion testing in its Industrial Scale Burner Facility (ISBF) at its U.S. Power Plant Laboratories facility in Windsor, Connecticut in support of these objectives. The NOx reduction approach was to optimize near-field combustion to ensure that minimum NOx emissions are achieved with minimal impact on unburned carbon in ash, slagging and fouling, corrosion, and flame stability/turn-down. Several iterations of CFD and combustion testing on a Midwest coal led to an optimized design, which was extensively combustion tested on a range of coals. The data from these tests were then used to validate system costs and benefits versus SCR. Three coals were evaluated during the bench-scale and large pilot-scale testing tasks. The three coals ranged from a very reactive subbituminous coal to a moderately reactive Western bituminous coal to a much less reactive Midwest bituminous coal. Bench-scale testing was comprised of standard ASTM properties evaluation, plus more detailed characterization of fuel properties through drop tube furnace testing and thermogravimetric analysis. Bench-scale characterization of the three test coals showed that both NOx emissions and combustion performance are a strong function of coal properties. The more reactive coals evolved more of their fuel bound nitrogen in the substoichiometric main burner zone than less reactive coal, resulting in the potential for lower NOx emissions. From a combustion point of view, the more reactive coals also showed lower carbon in ash and CO values than the less reactive coal at any given main burner zone stoichiometry. According to bench-scale results, the subbituminous coal was found to be the most amenable to both low NOx, and acceptably low combustibles in the flue gas, in an air staged low NOx system. The Midwest bituminous coal, by contrast, was

  19. Requirements, needs, and concepts for a new broad-application test reactor

    International Nuclear Information System (INIS)

    For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need

  20. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  1. Health physics experiences in the operation of Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    This paper presents the health physics experience gained with the operation of the Fast Breeder Test Reactor (FBTR), which was made critical in October 1985. Major operations that were carried out and the associated health physics surveillance are highlighted. (author)

  2. Research reactor fuel bundle design review by means of hydrodynamic testing

    International Nuclear Information System (INIS)

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author)

  3. Optimization of the sipping test procedure for WWER-type reactors

    International Nuclear Information System (INIS)

    The endeavour to provide for a lower activity level of the fission products in the power reactor coolant during operation is associated with timely identification and unloading of failed fuel elements from the core. In this connection studies have been performed to optimize the sipping test procedures existing for the WWER-440 and WWER-1000 type reactors. Possible incorrectness in the sipping test is discussed, some methods for increasing accuracy and reliability of the results are proposed. (orig.)

  4. CFD Analysis of NOx Emissions of a Natural Gas Lean Premixed Burner for Heavy Duty Gas Turbine

    OpenAIRE

    Andreini, A.; Cerutti, M; B. Facchini; Innocenti, A.

    2015-01-01

    The present work presents a numerical analysis of a low NOx partially premixed burner for heavy duty gas turbine. The first part of the paper is focused on the study of the premixing process inside the burner using standard RANS CFD approach. The resulting profiles at different test points have been used to perform reactive simulations of an experimental test rig, where exhaust NOx emissions were measured. A reliable numerical setup was found comparing predicted and measured NOx emissions at ...

  5. The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1979-03-01

    Recent U.S. Department of State action to restrict the shipment and use of highly enriched uranium for research and test reactors has renewed fuel development activity. The objective of these development activities is to increase the total uranium loading in the fuel meat so that enrichment reduction can be accomplished without significant performance penalties. This report characterizes the status and the potential for development of the currently utilized plate-type fuels for research and test reactors. The report also characterizes the newer high-density fuels which could be utilized in these reactors and indicates the impact of the utilization of both the new and current fuels on enrichment reduction.

  6. Proceedings of the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program

  7. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  8. Sensitivity Test of RIA in 5MW Research Reactor during Startup

    International Nuclear Information System (INIS)

    During startup operation, control absorber rods are not located in the equal critical position since they can be manually controlled by an operator without position limitation. Therefore, the power peaking factor in this control mode becomes larger due to the skewed power shape, making the accident consequence worse. In research reactor, the reactor protection system (RPS) has linear power trip and power lograte trip for a safe shutdown of reactor in the accident, and the occurrence of those trips depend both on the initial reactor power and the reactivity insertion rate. Therefore, with a series of sensitive analyses, we identified the most severe combination of initial conditions among the various initial reactor powers and reactivity insertion rates. The model reactor in this analysis is a 5MW pooltype research reactor having two different operation modes; a power operation and a training operation.. Since the accident occurs during startup of the reactor, the training mode without a forced convection results in more severe consequences in a view of fuel integrity. Therefore, the inadvertent withdrawal of a control rod during a startup of training operation is analyzed as a limiting case of the accident. Sensitivity tests with combinations of different initial reactor powers and reactivity insertion rates are performed for an inadvertent CAR withdrawal during startup of the training operation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-01

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

  10. FLICA III M - reactors or test loops thermohydraulic computer code

    International Nuclear Information System (INIS)

    The FLICA III M code issued from the FLICA III code, of which it is the present stage of the development. This program calculates the flow and their transfer in steady and transient state in complex geometry described by subchannels. It is particularly used for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. A new solution method for the hydraulic problem is developed. It gives short computer times and allows the calculation of large subchannels sets. This makes possible the detailed calculations of hot subchannels jointly with these of rod bundles set in powered reactor cores. The equations solved take into account all the significant terms of the fundamental thermal-hydraulic equations and present models for turbulence and two-phase flows. The solution method couples together all the physical variables and makes possible the detailed description of complex flows

  11. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  12. The Jules Horowitz reactor: an European material testing reactor main design options

    International Nuclear Information System (INIS)

    CEA has launched an initiative for the construction in Cadarache of a second generation MTR - Jules Horowitz Reactor JHR - To be operated as an international user facility addressing major industrial and societal needs for several decades ahead. - Providing effective experimental capabilities to study material and fuel under irradiation. - Supporting safety, reliability and improved economy of commercial reactors. - Assessing future systems required for sustainable power supply and/or waste management, including future fission or fusion systems and hybrid systems. - Helping national and international decision processes on for instance nuclear safety issues or matters related to the development of nuclear energy. - Supporting education and training for future nuclear scientists and engineers and inducing structuring effects in the international nuclear community. (author)

  13. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  14. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  15. Development and certification of the innovative pioneer oil burner for residential heating appliances

    Energy Technology Data Exchange (ETDEWEB)

    Kamath, B. [Heat Wise Inc., Ridge, NY (United States)

    1997-09-01

    The Pioneer burner represents another important milestone for the oil heat industry. It is the first practical burner design that is designated for use in small capacity heating appliances matching the needs of modern energy efficient home designs. Firing in the range of 0.3 GPH to 0.65 GPH (40,000-90,000 Btu/hr) it allows for new oil heating appliance designs to compete with the other major fuel choices in the small design load residential market. This market includes energy efficient single family houses, town-houses, condominiums, modular units, and mobile homes. The firing range also is wide enough to cover a large percentage of more conventional heating equipment and home designs as well. Having recently passed Underwriters Laboratory certification tests the burner in now being field tested in several homes and samples are being made available to interested boiler and furnace manufacturers for product development and application testing.

  16. Application of a Central Composite Design for the Study of NOx Emission Performance of a Low NOx Burner

    Directory of Open Access Journals (Sweden)

    Marcin Dutka

    2015-04-01

    Full Text Available In this study, the influence of various factors on nitrogen oxides (NOx emissions of a low NOx burner is investigated using a central composite design (CCD approach to an experimental matrix in order to show the applicability of design of experiments methodology to the combustion field. Four factors have been analyzed in terms of their impact on NOx formation: hydrogen fraction in the fuel (0%–15% mass fraction in hydrogen-enriched methane, amount of excess air (5%–30%, burner head position (20–25 mm from the burner throat and secondary fuel fraction provided to the burner (0%–6%. The measurements were performed at a constant thermal load equal to 25 kW (calculated based on lower heating value. Response surface methodology and CCD were used to develop a second-degree polynomial regression model of the burner NOx emissions. The significance of the tested factors over their respective ranges has been evaluated using the analysis of variance and by the consideration of the coefficients of the model equation. Results show that hydrogen addition to methane leads to increased NOx emissions in comparison to emissions from pure methane combustion. Hydrogen content in a fuel is the strongest factor affecting NOx emissions among all the factors tested. Lower NOx formation because of increased excess air was observed when the burner was fuelled by pure methane, but this effect diminished for hydrogen-rich fuel mixtures. NOx emissions were slightly reduced when the burner head was shifted closer to the burner outer tube, whereas a secondary fuel stream provided to the burner was found to have no impact on NOx emissions over the investigated range of factors.

  17. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  18. Sipping tests for the irradiated fuel elements of the TR-2 research reactor

    International Nuclear Information System (INIS)

    Sipping tests have been performed for fuel elements of the TR-2 reactor at Cekmece Nuclear Research and Training Center (CNRTC), in order to find out which one failed in the core. A sipping assembly has been constructed and placed in the pool of the TR-2 reactor. The assembly identifies leaking fuel elements by collecting and measuring 137Cs that leak out from the defective fuel elements. 31 fuel elements in the reactor have been tested for the clad integrity. The measured 137Cs activity of the fuel element with an identification number S-104 is a 10247 Bq/(0.3 l). This value is approximately 234 times greater than the average of the other tested fuel elements in the reactor. (orig.)

  19. Welding of sule elements for nuclear reactors with solid state YAG laser using instrumentated testing equipments

    International Nuclear Information System (INIS)

    The instrumentation of the equipment for carrying out safety tests on fuel elements for nuclear reactors requires special thermocouples adapted to the prevailing agressive medium. The investigations described deal essentially with the operational and metallurgical weldability tests out on the safety test zircaloy piping in the pressurized water circuit (PHEBUS-programme)

  20. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors (HPRR)) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  1. Manufacturing and material properties of forgings for the reactor pressure vessel of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    For the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) which has been developed by the Japan Atomic Energy Research Institute (JAERI), 2.25Cr-1Mo steel is used for the first time in the world for a nuclear reactor pressure vessel. The RPV is 13.2 m in height and 5.5 m in internal diameter. Operation temperature is about 400 C and the internal pressure is 4 MPa. A material confirmation test has been carried out to demonstrate good applicability of forged low Si 2.25Cr-1Mo steel to the RPV of the HTTR. Recently, Japan Steel Works has succeeded in manufacturing large size ring forgings and a large size forged cover dome integrated with nozzles for the stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for the stand pipe. (orig.)

  2. Thermophotovoltaic generation of electricity in a gas fired heater: Influence of radiant burner configurations and combustion processes

    International Nuclear Information System (INIS)

    With recent advances in low bandgap thermophotovoltaic (TPV) devices, further research into the radiant burner and its effect on the performance of TPV systems is particularly needed. The present work investigates various gas fired radiant burner/emitters and the influence of the combustion processes on radiant power and radiant efficiency. The performance tests with the burner/emitters have been conducted in a TPV self powered heater (mini cogenerator). It is shown that the radiant burner performance is affected markedly by the combustion parameters. Care must be taken to diminish the risk of flashback for the surface flame type burner. The maximum radiant power density and radiant efficiency of the burner/emitters have been determined. This is of great interest to TPV generation in gas fired heating appliances. Furthermore, the maximum electric power generated by the GaSb TPV converter is measured under a range of operating conditions for the different burner/emitter configurations. An electric power density of 0.332 W/cm2 has been achieved. Finally, the cogenerating aspects of the TPV systems are discussed

  3. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  4. HTGR actinide burner feasibility studies: Calculation scheme related considerations

    International Nuclear Information System (INIS)

    At the CEA, the actinides burner version of the prismatic block-type reactor is currently investigated, including studies about the design proposed by General Atomics. The purpose of this paper is essentially to evaluate the capability of the deterministic methods to calculate a wide range of core configurations. In the first part of the paper, the analysis is carried out on the 'Deep Burner' fuel element geometry. The fuel element calculations are performed with both Transport code APOLLO2 and Monte-Carlo code TRIPOLI4. This preliminary analysis shows the reliability of the deterministic code APOLLO2 to calculate heterogeneous fuel element configurations (fuel element loaded with plutonium and minor actinides). In the second part, the analysis deals with the core geometry in order to estimate the impact of some physical assumptions on the fine fuel isotopic depletion. Due to the strong spectrum transient in the core, it turns out that the transuranic mass balances in a GT-MHR cannot be estimated easily from fuel element calculations but rather need the use of a core modeling approach taking into account the presence of the graphite reflectors. Two different methods based on a fine core Diffusion calculation in CRONOS2 and a simplified Transport calculation in APOLLO2 are investigated in this paper. (authors)

  5. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  6. The heating operational summarization in three winters of a 5 MW test heating reactor

    International Nuclear Information System (INIS)

    The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

  7. Full conversion of materials and nuclear fuel in TRIGA SSR 14 MW research and test reactor

    International Nuclear Information System (INIS)

    During 1952-2005 General Atomics built and commissioned worldwide 62 TRIGA research reactors. Almost all reactors built by General Atomics use Low Enriched Uranium (19,9%). One exception was the TRIGA reactor from ICN Pitesti. The transition from HEU to LEU utilization, called a core conversion, is supported by Department of Energy - USA and Member States in the project 'Reduced Enrichment in Research and Testing Reactors (RERTR)' and by Member States and IAEA through Technical Cooperation programmes. The activities that are related to core conversion are managed and reviewed as refueling operations. This type of activities was performed at least six times from reactor commissioning stage until now. The implied personnel in this type of activities is licensed by CNCAN, the Romanian Regulatory Body and periodically trained. (authors)

  8. Destruction of weapons-grade plutonium with pebble bed type HTGRs using burner balls and breeder balls

    International Nuclear Information System (INIS)

    As the method of disposing the plutonium coming from disassembled weapons, the method of burning the fuel in which the plutonium is mixed with a parent material in LWRs or the disposal by glass solidification is proposed. In the former method, it is desirable to do the reprocessing of spent fuel for effectively utilizing fission products. The latter method needs watch against the diversion of the plutonium. The authors devised the method of effectively annihilating plutonium by separating into the burner balls of plutonium and the breeder balls of a parent material, and burning those by mixing in a pebble bed type high temperature gas-cooled reactor, while continuously exchanging them. It was clarified from the aspect of nuclear characteristics that by using this method, 239Pu can be annihilated to the state of enabling the direct abandonment without reprocessing. The flow of burner balls and breeder balls in the reactor is shown, and multi-pass fuel exchange method was adopted to burn Pu in burner balls up. The rate of Pu annihilation was determined by the change of the amount of Pu for the burnup evaluated by lattice burning calculation. The maximum amount of Pu charge in one burner ball is limited by the maximum allowable power output of burner balls. (K.I.)

  9. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  10. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  11. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  12. Fuel subassembly leak test chamber for a nuclear reactor

    International Nuclear Information System (INIS)

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained

  13. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  14. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  15. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: 'contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).' This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  16. IMPROVEMENT OF OPERATIONAL CHARACTERISTICS OF ELECTRIC COOKER BURNERS

    Directory of Open Access Journals (Sweden)

    I. M. Kirick

    2015-01-01

    Full Text Available On the basis of a complex theoretical and experimental investigations a principally new design of small inertial burner for electric cookers has been developed that significantly out-perform burners of conventional types. 

  17. Batch Tests To Determine Activity Distribution and Kinetic Parameters for Acetate Utilization in Expanded-Bed Anaerobic Reactors

    OpenAIRE

    Fox, Peter; Suidan, Makram T.

    1990-01-01

    Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distrib...

  18. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  19. Disposition of plutonium with HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade Plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. The power coefficient was made negative by reducing the parasitic neutron absorption reaction rate of 135Xe. (author)

  20. Performance and testing of refractory alloy clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO2) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at .% burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  1. Handbook of materials testing reactors and ancillary hot laboratories in the European Community

    International Nuclear Information System (INIS)

    The purpose of this Handbook is to make available to those interested in 'in-pile' irradiation experiments important data on Materials Testing Reactors in operation in the European Community. Only thermal reactors having a power output of more than 5 MW(th) are taken into consideration. In particular, detailed technical information is given on the experimental irradiation facilities of the reactors, their specialized irradiation devices (loops and instrumented capsules), and the associated hot cell facilities for post-irradiation examination of samples

  2. Test programme for the nuclear fuel of the Jules Horowitz reactor starts

    International Nuclear Information System (INIS)

    In Cadarache in the south of France, an advanced material test reactor is currently being built. The Jules Horowitz reactor is an initiative of the French Atomic Energy Commission ((CEA), supported by the European Commission and research centres in the Czech Republic, Spain, Finland and Belgium, plus a number of large energy companies. On the request of CEA, SCK-CEN started characterising the nuclear fuel for this reactor, in its own BR2. The article describes SCK-CEN's programme on the characterisation of new nuclear fuel elements.

  3. Influential parameters of nitrogen oxides emissions for microturbine swirl burner with pilot burner

    Directory of Open Access Journals (Sweden)

    Adžić Miroljub M.

    2010-01-01

    Full Text Available Swirl burners are the most common type of device in wide range of applications, including gas turbine combustors. Due to their characteristics, swirl flows are extensively used in combustion systems because they enable high energy conversion in small volume with good stabilization behavior over the wide operating range. The flow and mixing process generated by the swirl afford excellent flame stability and reduced NOx emissions. Experimental investigation of NOx emission of a purposely designed micro turbine gas burner with pilot burner is presented. Both burners are equipped with swirlers. Mixtures of air and fuel are introduced separately: through the inner swirler - primary mixture for pilot burner, and through the outer swirler - secondary mixture for main burner. The effects of swirl number variations for the both burners were investigated, including parametric variations of the thermal power and air coefficient. It was found that the outer swirler affects the emission of NOx only for the air coefficient less than 1.4. The increase of swirl number resulted in decrease of NOx emission. The inner swirler and thermal power were found to have negligible effect on emission.

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

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

    International Nuclear Information System (INIS)

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

  6. In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

    International Nuclear Information System (INIS)

    The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

  7. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  8. Isotope correlation studies relative to high enrichment test reactor fuels

    International Nuclear Information System (INIS)

    Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched 235U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched 235U fuel, the correlation of the isotopic ratio 143Nd/145+146Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The 137Cs/135Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum

  9. Siting evaluation of High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

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

  10. Application of the hybrid diffusion-transport spatial homogenization method to a high temperature test reactor benchmark problem

    International Nuclear Information System (INIS)

    The recently developed Hybrid Diffusion-Transport Spatial Homogenization (DTH) Method was previously tested on a benchmark problem typical of a boiling water reactor. In this paper, the DTH method is tested in a 1-D benchmark problem based on the Japanese High Temperature Test Reactor (HTTR). This acts as a verification of the method for a reactor that is optically thinner than the original BWR test benchmark. (author)

  11. Design characteristics and startup tests of HANARO. The newly in-service Korean research Reactor

    International Nuclear Information System (INIS)

    The Korea Atomic Energy Research Institute has successfully completed the commissioning tests, including the long-term operational performance test for the HANARO, a newly in-service multi-purpose research reactor in Korea. This paper presents the design characteristics of the HANARO and a brief description on the startup tests carried out at zero and low power during its commissioning. The reactor is now at the shutdown state for the periodic surveillance after cycle 1 operation and is waiting for the additional fuel loading to configure the cycle 2 core. The reactor is expected to undergo five more cycles of operation in this year. The important physics tests will be conducted at the beginning and at the end of each cycle. (author)

  12. DESIGN REPORT LOW-NOX BURNERS FOR PACKAGE BOILERS

    Science.gov (United States)

    The report describes a low-NOx burner design, presented for residual-oil-fired industrial boilers and boilers cofiring conventional fuels and nitrated hazardous wastes. The burner offers lower NOx emission levels for these applications than conventional commercial burners. The bu...

  13. DESIGN REPORT: LOW-NOX BURNERS FOR PACKAGE BOILERS

    Science.gov (United States)

    The report describes a low-NOx burner design, presented for residual-oil-fired industrial boilers and boilers cofiring conventional fuels and nitrated hazardous wastes. The burner offers lower NOx emission levels for these applications than conventional commercial burners. The bu...

  14. LASER-ENHANCED IONIZATION SPECTROMETRY WITH A TOTAL CONSUMPTION BURNER

    OpenAIRE

    Green, R; Hall, Janet

    1983-01-01

    This paper describes the use of a total consumption burner as an analytical atom reservoir for laser-enhanced ionization spectrometry. A total consumption burner and premixed burner are compared for limits of detection and matrix interferences. These results demonstrate that high sensitivity laser-enhanced ionization measurements are possible in adverse sample environments where traditional methods of optical spectrometry have proven inadequate.

  15. An evaluation of lifetime of JMTR and a study of new materials testing reactor

    International Nuclear Information System (INIS)

    The pressure vessel of the JMTR, which is made of austenite stainless steel, was designed so as to be in service for at least 20 years. Hence, the JMTR is expected to be in use until at least 1989, as the operation was started in 1969. But it was not clear how long the JMTR will be in use thereafter. Therefore, based on the ASME Section III, a review of the lifetime was made this time with the result that the pressure vessel can be in service for further 30 years hereafter. In evaluating the lifetime of the JMTR, the pressure vessel and its adaptors such as the grid plate are to be taken up and rather than these conponents are not to be taken up, as latters are exchangeable or repairable. A review of current requests for reactor irradiation was parallelly made, and a type of new reactor (JMTR-II is a tentative name) was preliminarily surveyed. Though being able to be in use for a fairly long time hereafter, as mentioned above, the JMTR has a possibility to be shutdown due to becomming stale like almost all research and test reactors. Therefore, it is not so early to survey the new reactor this time. New fields of reactor usage such as materials irradiation and tritium production for developing controlled thermonuclear reactor are to be considered in the design of the JMTR-II. Some requests from various fields for new reactor are incompatible, hence compromise is to be inevitable, because there might be no possibility to construct a couple of research and test reactors in Japan at a time. A light water moderated and cooled, pressurized type is selected as a recommendable candidate of the JMTR-II, after comparing some reactor types considered, and nuclear survey calculations were performed on it. (author)

  16. 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

    2012-01-01

    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...

  17. Seismic hazard analysis for the NTS spent reactor fuel test site

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, K.W.

    1980-05-02

    An experiment is being directed at the Nevada Test Site to test the feasibility for storage of spent fuel from nuclear reactors in geologic media. As part of this project, an analysis of the earthquake hazard was prepared. This report presents the results of this seismic hazard assessment. Two distinct components of the seismic hazard were addressed: vibratory ground motion and surface displacement. (ACR)

  18. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  19. Thermohydraulic tests in the area of reactor safety done in CDTN

    International Nuclear Information System (INIS)

    The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor)

  20. Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    K. L. Davis; D. L. Knudson; J. L. Rempe; J. C. Crepeau; S. Solstad

    2015-07-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.