WorldWideScience

Sample records for advanced fissile fuel

  1. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U3O8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U3O8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  2. Irradiation behaviour of coated fuel particles for the fissile/fertile particle system

    International Nuclear Information System (INIS)

    The cross evaluation of 30 irradiation experiments which were carried out in the last ten years in order to test fuel particles for the separate use of high enriched uranium in fissile particles and thorium in fertile particles, led to the following results: An oxide-based fissile/fertile particle system (UO2 fissile kernel/ThO2 fertile kernel) can be used as well as the American carbide/oxide particle system (UC2 fissile kernel/ThO2 fertile kernel) under the operation conditions of a high temperature reactor with spherical fuel elements. The swelling of fissile kernels as a consequence of fission gas pores is much more pronounced in UO2 than in UC2 fissile kernels but the buffer layer copes with the swelling without any problems. Ceramic kernel additives (e.g. Al2O3)as well as carbon additives proved not to be suitable because they deteriorate the mechanical properties of the fissile kernels. Kernel migration in a temperature gradient ('amoeba effect') is observed during irradiation of UO2 fissile particles but this does not cause any coating failure. The amoeba effect is suppressed completely by 10% UC2 additives to the UO2 kernel. The silicon carbide interlayer is absolutely necessary for an efficient retention of the solid fission products and has also proved successfully for fertile particles. A measurable fraction of defective particle coatings was not observed before exceeding the target values of burnup and fast neutron fluence. The irradiation-induced dimensional changes of the graphite matrix are independent of the fuel volume loading and have no influence on the irradiation behaviour of the embedded fissile and fertile particles. (orig./IHOE)

  3. Development of NDA reactivity measurement method to determine the fissile contents for spent fuel

    International Nuclear Information System (INIS)

    NDA reactivity measurement method was developed and simulated for the possible use in determination of fissile content and its reactivity characteristics of spent fuel material. This is necessary and practical for the purpose of safeguards implementation and timely control of process line under the condition of very high radiation environment. The neutron detector response due to induced fission from fissile material in spent fuel is calculated by using the MCNP model. The effects of detector response on fissile content, burnup and initial enrichment of fuel are investigated. It shows that the detector response of spent fuel contained 239Pu and fission products is slightly higher than that of flesh fuel with coresponding 235U enrichment

  4. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Levent, E-mail: levent.ozdemir@taek.gov.tr [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey); Acar, Banu Bulut; Zabunoglu, Okan H. [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey)

    2011-02-15

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of {sup 239}Pu and {sup 241}Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  5. Optimisation Studies of Accelerator Driven Fertile to Fissile Conversion Rates in Thorium Fuel Cycle

    OpenAIRE

    Bungau, Cristian; Barlow, Roger; Cywinski, R.

    2012-01-01

    The need for proliferation-resistance, longer fuel cycles, higher burn up and improvedwaste form characteristics has led to a renewed worldwide interest in thorium-based fuels and fuel cycles. In this paper the GEANT4 Monte Carlo code has been used to simulate the Thorium-Uranium fuel cycle. The accelerator driven fertile to fissile conversion rates have been calculated for various geometries. Several new classes have been added by the authors to the GEANT4 simulation ...

  6. Neutron multiplication method for measuring the amount of fissile isotopes in the spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Pytel, K. [Institute of Atomic Energy, 05-400 Otwock-Swierk (Poland); Abou-Zaid, A.A. [Atomic Energy Authority, 13759 Cairo (Egypt)

    2001-07-01

    A nondestructive assay method for determination the amount of fissile materials content along the vertical axis of irradiated fuel is presented. The method, called neutron multiplication method, can be realized as passive measurement technique and the active one. The Monte Carlo code has been used for the neutron transport simulation and optimization of the measuring equipment geometry. On the basis of these results, a preliminary experimental stand for MARIA reactor fuel investigation has been designed and the measurements have been performed for the fresh fuel and the fuel mock-up. Based upon both numerical and experimental simulations, an ultimate measuring stand has been designed and the measurements for MARIA spent fuel assemblies as well as for the fresh fuel and mock-up of the fuel have been carried out. The results showed that the active neutron technique does not provide sufficient resolution of the distribution of the amount of fissile materials. But rather can be applied for measurement of the absolute value. The passive one can be used to restore the distribution of the bum-up and the amount of fissile materials along the axial length of the spent fuel assembly. (author)

  7. The neutron emission method for determination of fissile materials within the spent fuel equipment optimization

    Energy Technology Data Exchange (ETDEWEB)

    Abou-Zaid, A. [Nuclear Research Center, Atomic Energy Authority, 13759- Cairo (Ethiopia); Pytel, K. [Atomic Energy Institute, Research Reactor Center, 05-400 Otwock-Swierk (Poland)

    1998-07-01

    A nondestructive assay method using neutron technique for determination of the fissile isotopes content along the irradiated fuel rods of MARIA reactor is presented. This method is based on detection of the fission neutrons emitted from external neutron source and multiplied by the fissile isotopes U-235, Pu-239, and Pu-241 within the fuel rod. Neutrons emitted from the spent fuel originate mainly from induced fission in the fissile material and source neutrons penetrating the fuel rod without interaction. Additionally, the neutrons from ({alpha}, n) reaction and spontaneous fission of actinide isotopes contribute in the total population of emitted ones. The method gives a chance to perform an experimental calibration of the equipment using two points: fresh fuel rod (maximum signal plus background) and its mock-up (background). The Monte Carlo code has been used for the geometrical simulation and optimization of the measuring equipment: neutron source, moderating container, collimator, and the neutron detector. The results of the calculation show that the moderating container of 30 cm length and 32 cm diameter and a collimator of 26 cm length, 6.8 cm width, and 2 cm height are the optimal configuration. With respect to the fission chamber position, the number of neutrons has been calculated as a function of distance from the fuel rod surface in the case of fresh fuel and its mock-up. The distance, at which the ratio of the signal to background has its maximum, has been found at 4.5 cm far from the outer surface of the fuel. (author)

  8. On international criticality codes for fuel pellets in fissile solution

    International Nuclear Information System (INIS)

    The reference calculations, based on the APOLLO-Pic method implemented in the framework of this study, demonstrated that the actual reactivity variation (benchmark n0 20) is a monotonic decrease with pellet dissolution. At the opposite of the contributor's results, based on the international criticality code SCALE, the reactivity loss with dissolution is weak. The discrepancy is mainly due to 238U resonant absorption which can induce, in this fuel double heterogeneity problem n0 20, as much as -30 000 pcm K∞ underestimation. It was pointed out that design-oriented transport codes must be improved by accurate deterministic formalisms: PIC equivalence method, subgroup theory (WIMSE), ultrafine slowing-down calculation (ROLAIDS). Ultimate confirmation of the reference results presented in this paper should be provided by a set of critical experiments which mock-up hypothetical dissolver geometries. Finally it should be noted that thanks to the interest and the efforts of the OECD/NEA Criticality Working Group in performing the international benchmark exercise and in pursuing the explanation of the discrepancies, a potentially dangerous inadequacy in criticality calculation methods was exposed and resolved

  9. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  10. Determination of dependence of fissile fraction in MOX fuels on spent fuel storage period for different burnup values

    International Nuclear Information System (INIS)

    Highlights: ► In a previous study, an expression to calculate fissile fraction of MOX for various burnups was obtained for 5-year cooled SF. ► In this follow-up study, a correction factor for spent fuel storage periods other than 5 years is derived. ► Thus, one major restriction on use of the expression derived in the initial study is eliminated. - Abstract: The purpose of this technical note is to remove one of the limitations of a derived expression in a previously published article (Özdemir et al., 2011). The original article focused on deriving (computationally) an expression for calculating total fissile fraction of mixed oxid (MOX) fuels depending on discharge burnup of spent fuel and desired burnup of MOX fuel; consequently, such an expression was obtained and put forward, together with its limitations. One of the limitations has been that all the computations and therefore the resulting expression are based on the assumption of a spent fuel storage period of 5 years. This follow-up study simply aims to obtain a correction factor for spent fuel storage periods other than 5 years; thus to remove one major restriction on use of the expression derived in the original article

  11. Fissile compound - Inert matrix compatibility studies for the development of gas cooled fast reactor fuels

    International Nuclear Information System (INIS)

    Helium-Cooled High-Temperature Fast Reactors have a high potential for transmutation of minor actinides (Pu, Am, Cm... ). In this kind of reactor, the fuel temperature would be 1200 deg C in use and the inert matrix should retain the fission products in the fuel structure up to 1600 deg C. The fissile compound would be (U,Pu)C or (U,Pu)N owing to their high density, good thermal conductivity and refractory behavior. SiC, TiC, ZrC and TiN, ZrN would be the inert matrix surrounding (U,Pu)C or (U,Pu)N fissile compounds. This study is devoted to the chemical compatibility between UC or UN and inert matrix in the 1200 deg C - 2000 deg C temperature range. In order to achieve a limited number of specific experiments, thermodynamic calculations are realized using the thermodynamic data provided either by the Thermodata database or from the literature. (authors)

  12. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Science.gov (United States)

    Lee, Tae-Hoon; Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J.

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the 244Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective 239Pu mass was introduced by weighing the relative contribution to the signal of 235U and 241Pu compared to 239Pu and the calibration curves of DDA count rate vs. 239Pu eff were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10 9 n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  13. Development of the fuel of gas fast future reactors. Study of the compatibility between fissile compound and inert matrix

    International Nuclear Information System (INIS)

    The gas coolant fast future reactors require to develop new types of fuels. The fissile compounds (U, Pu, AM)C and (U, Pu, AM)N are retained, on account of their high density, their refractory character and their good thermal conductivity. The choice for the inert matrix, which will contain the fissile compound, bears on ceramics: SiC, TiC and ZrC in the case of actinides monocarbide and SiC, TiN and ZrN in the case of actinides mono-nitrides. The chemical compatibility between the main fissile compound (UN or UC) and the inert matrices SiC and TiN has been studied in the range of 1200 C (nominal running conditions) and 2000 C (accidental running conditions). The thermodynamic compatibility calculations allow to determine particularly at which temperature a liquid phase risks to appear by reaction between the fissile compound and the inert matrix and to limit the number of tests to carry out. This approach is completed by specific interaction tests carried out under different atmospheres. (O.M.)

  14. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  15. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae-Hoon, E-mail: typhoon@kaeri.re.kr [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Korea Atomic Energy Research Institute, 150-1 Dukjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighing the relative contribution to the signal of {sup 235}U and {sup 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  16. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  17. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  18. LIGHTBRIDGE corporation advanced metallic fuel

    International Nuclear Information System (INIS)

    Lightbridge Corporation is developing a metallic nuclear fuel which utilizes an innovative fuel rod geometry and composition to improve power plant economics and enhance the performance and safety of commercial light water reactors. The versatile fuel can utilize uranium or plutonium as the fissile component. The fuel is fully compatible with existing light water reactor designs and requires no major changes to reactor operations. The metallic fuel provides a durable solution that is also capable of operating at higher power density than existing fuels allowing for increased power output and cycle length compared to conventional oxide fuels. Lightbridge patented nuclear fuel technologies are designed to significantly enhance nuclear power industry economics and increase power output by: 1) extending fuel cycle length to 24 months or longer while simultaneously increasing power output by 10% or increasing power output by up to 17% with 18-month fuel cycles in existing pressurized water reactors (PWRs); 2) enabling increased reactor power output of up to 30% without changing core size in new build PWRs; and 3) reducing the volume of used fuel per kilowatt-hour as well as enhancing proliferation resistance of spent fuel. (author)

  19. Irradiation performance of (Th,U)02 fuel designed for advanced fuel cycle applications

    International Nuclear Information System (INIS)

    The reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impregnated fuel has operated to 650 MW.h/kg H.E. at 50-60 KW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 MW.h/kg H.E. at a maximum linear power of 60 KW/m. The latter two experiments have operated without defects and with fission gas release less than that for U02 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale. We will continue monitoring the extruded fuel to a burnup approaching 1000 MW.h/kg H.E., as an indicator for the performance expected from co-precipitated (Th,U)02 or mechanically-mixed (Th,U)02 with good fissile homogeneity

  20. Advanced coated particle fuels

    International Nuclear Information System (INIS)

    The coated particle fuel (cpf) has been developed for use in high-temperature gas-cooled reactors, but it may find applications in other types of reactors. In JAERI, besides the development of cpf for High Temperature Engineering Test Reactor, conceptual studies of the cpf applications in actinide burner reactors and space reactors have been made. The conceptual design studies as well as the research and development of advanced coatings, ZrC and TiN, are reviewed. (author)

  1. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  2. Advanced Fuels Campaign 2012 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  3. Advanced fuel developments to improve fuel cycle cost in PWR

    International Nuclear Information System (INIS)

    Increasingly lower fuel cycle costs and higher plant availability factors have been two crucial components in keeping the overall cost of electricity produced by nuclear low and competitive with respect to other energy sources. The continuous quest to reduce fuel cycle cost has resulted in some consolidated trends in LWR fuel management schemes: smaller number of feed fuel assemblies with longer residence time; longer cycles, with 18-month cycle as the predominant option, and some plants already operating on, or considering, 24-month refueling intervals; higher power ratings with many plants undergoing power uprates. In order to maintain or improve fuel utilization for the longer cycles and/or higher power ratings, the licensed limits in fuel fissile content (5.0 w/o U235 enrichment) and discharge burnup (62 GWd/tHM for the peak pin) have been approached. In addition, Zr-based fuel cladding materials are also being challenged by the resulting increased duty. For the above reasons further improvements in fuel cycle cost have to overcome one or more of the current limits. This paper discusses an option to break through this 'stalemate', i.e. uranium nitride (UN) fuel with SiC clad. In UN the higher density of the nitride with respect to the oxide fuel leads to higher fissile content and reduction in the number of feed assemblies, improved fuel utilization and potentially higher specific powers. The SiC clad, among other benefits, enables higher clad irradiation, thereby exploiting the full potential of UN fuel. An alternative to employing UN fuel is to maintain UO2 fuel but boost the fissile content increasing the U235 enrichment beyond the 5 w/o limit. The paper describes and compares the potential benefits on fuel cycle cost of either option using realistic full-core calculations and ensuing economic analysis performed using Westinghouse in-house reactor physics tools and methodologies. (author)

  4. On-Going Comparison of Advanced Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Ralph G. Bennett; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mark Roth; J. D. Smith; Robert Hill; James Laidler; Kemal Pasamehmetoglu

    2004-10-01

    The Advanced Fuel Cycle Initiative (AFCI) program is addressing key issues associated with critical national needs. This paper compares the major options with these major “outcome” objectives - waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety as well as “process” objectives associated with readiness to proceed and adaptability and robustness in the face of uncertainties. Working together, separation, transmutation, and fuel technologies provide complete energy systems that can improve waste management compared to the current “once-through/no separation” approach. Future work will further increase confidence in potential solutions, optimize solutions for the mixtures of objectives, and develop attractive development and deployment paths for selected options. This will allow the nation to address nearer-term issues such as avoiding the need for additional geological repositories while making nuclear energy a more sustainable energy option for the long-term. While the Generation IV Initiative is exploring multiple reactor options for future nuclear energy for both electricity generation and additional applications, the AFCI is assessing fuel cycles options for either a continuation or expansion of nuclear energy in the United States. This report compares strategies and technology options for managing the associated spent fuel. There are four major potential strategies, as follows: · The current U.S. strategy is once through: standard nuclear power plants, standard fuel burnup, direct geological disposal of spent fuel. Variants include higher burnup fuels in water-cooled power plants, once-through gas-cooled power plants, and separation (without recycling) of spent fuel to reduce the number and cost of geological waste packages. · The second strategy is thermal recycle, recycling some fuel components in thermal reactors. This strategy extends the useful life of

  5. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  6. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  7. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  8. ARPA advanced fuel cell development

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, L.H.

    1995-08-01

    Fuel cell technology is currently being developed at the Advanced Research Projects Agency (ARPA) for several Department of Defense applications where its inherent advantages such as environmental compatibility, high efficiency, and low noise and vibration are overwhelmingly important. These applications range from man-portable power systems of only a few watts output (e.g., for microclimate cooling and as direct battery replacements) to multimegawatt fixed base systems. The ultimate goal of the ARPA program is to develop an efficient, low-temperature fuel cell power system that operates directly on a military logistics fuel (e.g., DF-2 or JP-8). The absence of a fuel reformer will reduce the size, weight, cost, and complexity of such a unit as well as increase its reliability. In order to reach this goal, ARPA is taking a two-fold, intermediate time-frame approach to: (1) develop a viable, low-temperature proton exchange membrane (PEM) fuel cell that operates directly on a simple hydrocarbon fuel (e.g., methanol or trimethoxymethane) and (2) demonstrate a thermally integrated fuel processor/fuel cell power system operating on a military logistics fuel. This latter program involves solid oxide (SOFC), molten carbonate (MCFC), and phosphoric acid (PAFC) fuel cell technologies and concentrates on the development of efficient fuel processors, impurity scrubbers, and systems integration. A complementary program to develop high performance, light weight H{sub 2}/air PEM and SOFC fuel cell stacks is also underway. Several recent successes of these programs will be highlighted.

  9. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  10. Advanced thermally stable jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  11. Advanced fuel chemistry for advanced engines.

    Energy Technology Data Exchange (ETDEWEB)

    Taatjes, Craig A.; Jusinski, Leonard E.; Zador, Judit; Fernandes, Ravi X.; Miller, James A.

    2009-09-01

    Autoignition chemistry is central to predictive modeling of many advanced engine designs that combine high efficiency and low inherent pollutant emissions. This chemistry, and especially its pressure dependence, is poorly known for fuels derived from heavy petroleum and for biofuels, both of which are becoming increasingly prominent in the nation's fuel stream. We have investigated the pressure dependence of key ignition reactions for a series of molecules representative of non-traditional and alternative fuels. These investigations combined experimental characterization of hydroxyl radical production in well-controlled photolytically initiated oxidation and a hybrid modeling strategy that linked detailed quantum chemistry and computational kinetics of critical reactions with rate-equation models of the global chemical system. Comprehensive mechanisms for autoignition generally ignore the pressure dependence of branching fractions in the important alkyl + O{sub 2} reaction systems; however we have demonstrated that pressure-dependent 'formally direct' pathways persist at in-cylinder pressures.

  12. ADVANCED FUELS CAMPAIGN 2013 ACCOMPLISHMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-10-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. Accomplishments made during fiscal year (FY) 2013 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section.

  13. Advanced Fuels Campaign Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kemal Pasamehmetoglu

    2011-09-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the 'Grand Challenge' for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  14. Advanced Fuels Campaign Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kemal Pasamehmetoglu

    2010-10-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the “Grand Challenge” for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  15. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  16. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  17. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  18. Accelerating fissile fuel breeding in FBR with natural safety features%加速增产核燃料的天然安全“核热泉”快中子增殖堆

    Institute of Scientific and Technical Information of China (English)

    吕应中

    2012-01-01

    为保证21世纪中国经济的持续稳定地高速增长,必须充分发挥核能的巨大潜力,使之配合其他可再生能源同步增长,及早大规模替代煤炭等化石能源.由于目前国内大量兴建的核电站以压水堆为主,需要消费大量天然铀资源,倚靠廉价铀供应难于维持长期增长,必须依靠快中子增殖生产人造裂变燃料——钚,才能摆脱天然铀原料短缺的束缚.然而,传统的快中子增殖堆的核燃料增产速度较慢,难于配合中国核电的高速增长.本文介绍一种先进快中于增殖堆(AFBR)方案,其中利用在线连续换料的空心球形燃料元件,依靠载热剂的出人口之间的温度差实现满功率自然循环,可以成倍地提高燃料比功率与核燃料增殖速度.本快中子增殖堆改进了俄罗斯称为“天然安全”的BREST铅冷快堆设计方案,成为无须人为控制的“核热泉”,它能在不设置加压泵及高位铅池的情况下,自动按外部负荷需要供应必要的热量,完全依靠自然循环将全部裂变热能及停堆后堆芯余热散出,不至对环境产生放射性污染.%To guarantee the rapid growth of the Chinese economy in 21" century, nuclear energy should be fully exploited, together with other renewable energies to replace coal and other depletive fossil fuels. Unfortunately, the major Chinese nuclear power plants under construction are mostly PWRs that would consume a lot of natural uranium during their operations. The availability of cheap natural uranium could seriously constraint the Chinese nuclear power development, unless artificial fissile fuel-plutonium is supplied from fast breeder reactors. The fissile nuclei production rate in the traditional fast breeders, however, seems too slow to match the rapid growth of nuclear power. A concept of the advanced fast breeder (AFBR) is introduced, therefore, to greatly accelerating the fissile fuel breeding process. In said breeder, the spherical hollow

  19. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  20. Nuclear propulsion technology advanced fuels technology

    Science.gov (United States)

    Stark, Walter A., Jr.

    1993-01-01

    Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

  1. Zirconia-magnesia inert matrix fuel and waste form: Synthesis, characterization and chemical performance in an advanced fuel cycle

    Science.gov (United States)

    Holliday, Kiel Steven

    There is a significant buildup in plutonium stockpiles throughout the world, because of spent nuclear fuel and the dismantling of weapons. The radiotoxicity of this material and proliferation risk has led to a desire for destroying excess plutonium. To do this effectively, it must be fissioned in a reactor as part of a uranium free fuel to eliminate the generation of more plutonium. This requires an inert matrix to volumetrically dilute the fissile plutonium. Zirconia-magnesia dual phase ceramic has been demonstrated to be a favorable material for this task. It is neutron transparent, zirconia is chemically robust, magnesia has good thermal conductivity and the ceramic has been calculated to conform to current economic and safety standards. This dissertation contributes to the knowledge of zirconia-magnesia as an inert matrix fuel to establish behavior of the material containing a fissile component. First, the zirconia-magnesia inert matrix is synthesized in a dual phase ceramic containing a fissile component and a burnable poison. The chemical constitution of the ceramic is then determined. Next, the material performance is assessed under conditions relevant to an advanced fuel cycle. Reactor conditions were assessed with high temperature, high pressure water. Various acid solutions were used in an effort to dissolve the material for reprocessing. The ceramic was also tested as a waste form under environmental conditions, should it go directly to a repository as a spent fuel. The applicability of zirconia-magnesia as an inert matrix fuel and waste form was tested and found to be a promising material for such applications.

  2. Neutron emission tomography for nuclear fissile materials safeguards

    International Nuclear Information System (INIS)

    Any nondestructive method for fissile assay in spent nuclear fuel must sample a substantial portion of the total number of reactor fuel elements to account for the radial and axial asymmetry of fissile material distribution that may arise during fuel burnup. Cross-sectional scans of the spatial distribution of the fissile isotopes can be tomographically reconstructed while a spent fuel element is being assayed in a lead slowing-down-time (SDT) spectrometer by employing image reconstruction techniques. The SDT approach employs threshold detectors (fission chambers), which measure the induced prompt fast fission neutrons emitted from the fissile isotopes in spent fuel while being interrogated with a slowing pulse of source neutrons. The objectives of this work were to examine to what extent a cross section representing an array of partially void (missing pins) and intact fuel pins could be reconstructed and how sensitive this method is to diversion, spatial burnup distribution, and gross fissile contents consistency

  3. Advanced Thermally Stable Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    A. Boehman; C. Song; H. H. Schobert; M. M. Coleman; P. G. Hatcher; S. Eser

    1998-01-01

    The Penn State program in advanced thermally stable jet fuels has five components: 1) development of mechanisms of degradation and solids formation; 2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles during thermal stressing; 3) characterization of carbonaceous deposits by various instrumental and microscopic methods; 4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and 5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics from coal.

  4. 49 CFR 173.420 - Uranium hexafluoride (fissile, fissile excepted and non-fissile).

    Science.gov (United States)

    2010-10-01

    ... uranium hexafluoride; and (iii) withstand the test specified in 10 CFR 71.73(c)(4) without rupture of the... 49 Transportation 2 2010-10-01 2010-10-01 false Uranium hexafluoride (fissile, fissile excepted....420 Uranium hexafluoride (fissile, fissile excepted and non-fissile). (a) In addition to any...

  5. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

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

  6. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

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

  7. Advanced Fuels Campaign FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  8. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238U + ∼0.7% 235U) and natural thorium (∼100% 232Th) resources. Currently, most power and research reactors use 235U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238U and 232Th are transmuted in the reactor to human-made 239Pu and 233U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238U and 235U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO2, in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels for sodium

  9. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities.

  10. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities. PMID:20223608

  11. Chemical Kinetic Modeling of Advanced Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    PItz, W J; Westbrook, C K; Herbinet, O

    2009-01-20

    Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.

  12. Uncertainty Analyses of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

    2008-12-12

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

  13. Uncertainty Analyses of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development

  14. LANL's Role in the U.S. Fissile Material Disposition Program

    Energy Technology Data Exchange (ETDEWEB)

    Whitworth, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kay, Virginia [NA-233

    2015-02-18

    The process of Fissile Material Disposition is in part a result of the Advanced Recovery and Integrated Extraction System (ARIES), which is an agreement between the U.S. and Russia to dispose of excess plutonium used to make weapons. LANL is one sight that aides in the process of dismantling, storage and repurposing of the plutonium gathered from dismantled weapons. Some uses for the repurposed plutonium is fuel for commercial nuclear reactors which will provide energy for citizens.

  15. Fuels for Advanced Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Fuels for advanced nuclear reactors differ greatly from conventional light water reactor fuels and vary widely between the different concepts, due differences in reactor architecture and deployment. Functional requirements of all fuel designs include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintaining a coolable geometry. In all cases, the anticipated fuel performance under normal or off-normal conditions is the limiting factor in reactor system design, and cumulative effects of increased exposure to higher burnup degrades fuel performance. In high-temperature (thermal) gas reactor systems, fuel particles of uranium dioxide or uranium oxycarbide particles are coated with layers of carbon and SiC (or ZrC). Such fuels have been used successfully to very high burnup (10-20% of heavy-metal atoms) and can withstand transient temperatures up to 1600 C. Oxide (pellet-type) and metal (pin-type) fuels clad in stainless steel tubes have been successfully used in liquid metal cooled fast reactors, attaining burnup of 20% or more of heavy-metal atoms. Those fuel designs are being adapted for actinide management missions, requiring greater contents of minor actinides (e.g. Am, Np, Cm). The current status of each fuel system is reviewed and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed

  16. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  17. Advanced Fuel Cycle Economic Sensitivity Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Shropshire; Kent Williams; J.D. Smith; Brent Boore

    2006-12-01

    A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

  18. Advanced Fuels Campaign FY 2011 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-11-01

    One of the major research and development (R&D) areas under the Fuel Cycle Research and Development (FCRD) program is advanced fuels development. The Advanced Fuels Campaign (AFC) has the responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. Accomplishments made during fiscal year (FY 20) 2011 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section. The order of the accomplishments in this report is consistent with the AFC work breakdown structure (WBS).

  19. Physics challenges for advanced fuel cycle assessment

    Energy Technology Data Exchange (ETDEWEB)

    Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti

    2014-06-01

    Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.

  20. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

    2002-07-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  1. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  2. Advanced Fuels Campaign FY 2010 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2010-12-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) Accomplishment Report documents the high-level research and development results achieved in fiscal year 2010. The AFC program has been given responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. The science-based approach combines theory, experiments, and multi-scale modeling and simulation aimed at a fundamental understanding of the fuel fabrication processes and fuel and clad performance under irradiation. The scope of the AFC includes evaluation and development of multiple fuel forms to support the three fuel cycle options described in the Sustainable Fuel Cycle Implementation Plan4: Once-Through Cycle, Modified-Open Cycle, and Continuous Recycle. The word “fuel” is used generically to include fuels, targets, and their associated cladding materials. This document includes a brief overview of the management and integration activities; but is primarily focused on the technical accomplishments for FY-10. Each technical section provides a high level overview of the activity, results, technical points of contact, and applicable references.

  3. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  4. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  5. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  6. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  7. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

  8. Modeling of advanced fossil fuel power plants

    Science.gov (United States)

    Zabihian, Farshid

    The first part of this thesis deals with greenhouse gas (GHG) emissions from fossil fuel-fired power stations. The GHG emission estimation from fossil fuel power generation industry signifies that emissions from this industry can be significantly reduced by fuel switching and adaption of advanced power generation technologies. In the second part of the thesis, steady-state models of some of the advanced fossil fuel power generation technologies are presented. The impacts of various parameters on the solid oxide fuel cell (SOFC) overpotentials and outputs are investigated. The detail analyses of operation of the hybrid SOFC-gas turbine (GT) cycle when fuelled with methane and syngas demonstrate that the efficiencies of the cycles with and without anode exhaust recirculation are close, but the specific power of the former is much higher. The parametric analysis of the performance of the hybrid SOFC-GT cycle indicates that increasing the system operating pressure and SOFC operating temperature and fuel utilization factor improves cycle efficiency, but the effects of the increasing SOFC current density and turbine inlet temperature are not favourable. The analysis of the operation of the system when fuelled with a wide range of fuel types demonstrates that the hybrid SOFC-GT cycle efficiency can be between 59% and 75%, depending on the inlet fuel type. Then, the system performance is investigated when methane as a reference fuel is replaced with various species that can be found in the fuel, i.e., H2, CO2, CO, and N 2. The results point out that influence of various species can be significant and different for each case. The experimental and numerical analyses of a biodiesel fuelled micro gas turbine indicate that fuel switching from petrodiesel to biodiesel can influence operational parameters of the system. The modeling results of gas turbine-based power plants signify that relatively simple models can predict plant performance with acceptable accuracy. The unique

  9. 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

  10. Economic projection for advanced fuel fabrication

    International Nuclear Information System (INIS)

    This paper presents a fabrication cost evaluation based on the status of MTR fuel development at NUKEM as of November 1980, and includes advanced chemical and mechanical processes starting with UF6 (including chemical scrap recovery). It also assumes that there will be just one fabrication line for each fuel type. UAlx-Al fuel with a uranium density of 1.0 g/cm3 was taken as the 100% cost reference. Excluding enrichment for the moment, the 100% reference values for all fuels are individually different for each reactor because of the different fuel element designs. Plate type, thickness of meat, length of meat, number of plates per element, bent or flat plates/tubes, upper and lower components, assembly design, materials, element quantity per order, etc., are all important in determining the final costs. New activity in research and development (R and D) has been started at NUKEM. In the very beginning, all the steps involved in fabrication were examined, starting with UF6 conversion and chemical treatment of the fuel (mainly, the oxide fuels), and continuing through powder production and plate fabrication, including intermediate chemical scrap recovery and final scrap recovery. The latter is very important because these procedures can also be used in reprocessing later on. If the mechanical or technical capability and/or the costs no longer makes sense, that R and D activity is stopped if there is an alternative fuel. At the moment, the prospects for alternative fuels are promising. The R and D on a particular fuel is stopped if its cost to the reactor operator would be increased by more than 30% over the current fuel cost. For this reason, the R and D on UAl alloy fuel was stopped about two years ago at a uranium density of roughly 1.2 g/cm3. Similarly, the R and D on UAIx fuel is now being stopped at around 2.2 g U/cm3 because the 30% limit has been reached. The R and D on U3O8 fuel is continuing up to about 3.2 g U/cm3, but there is some possibility of achieving

  11. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  12. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10-4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  13. Practice and prospect of advanced fuel management and fuel technology application in PWR in China

    International Nuclear Information System (INIS)

    Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

  14. Advanced technologies for power and fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Watts, J.U.; Mann, A.N. [US Department of Energy/National Energy Technology Lab., Pittsburgh, PA (United States)

    2001-07-01

    The Clean Coal Technology Program (CCT) being conducted by the United States Department of Energy (DOE) is a government and industry co-funded effort. The program's purpose is to demonstrate new generation of innovative, environmentally friendly processes that enhance the utilization of coal to meet increasing demand for electric power and fuels. Program demonstration areas include environmental control, advanced power generation, fuels production, and industrial applications. The CCT Program has now grown to maturity, with over 50% of the projects selected having successfully completed their demonstration goals and objectives. Under the CCT Program, nine advanced electric power generation projects and five coal processing for clean fuels projects were selected for full scale commercial demonstration. This paper provides the status, accomplishments and results of the most widely accepted technologies currently being commercialized under these two categories. The projects are (1) Atmospheric Fluidized-Bed Combustion (AFBC) at Jacksonville Electric Authority; (2) Integrated Gasification Combined-cycle (IGCC) at Wabash River, Tampa Electric and Kentucky Pioneer; and (3) Eastman Chemical's production of methanol via coal gasification using the LPMEOH{trademark} process. 7 figs., 7 tabs.

  15. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  16. Advanced fuel technology - A UK perspective

    International Nuclear Information System (INIS)

    The nuclear power industry in the United Kingdom is perhaps more diverse than in any other country. The diversity in design of stations is matched by a diversity in operating responsibility. The SGHWR and PFR are operated by the United Kingdom Atomic Energy Authority (UKAEA), 2 of the Magnox stations are owned and run by BNFL, 2 of the AGR stations and 1 Magnox station are controlled by the South of Scotland Electricity Board (SSEB), and the remaining reactors (including the Sizewell 'B' PWR) currently come under the responsibility of the Central Electricity Generating Board (CEGB) but will pass into the control of a new state-run company when the rest of the CEGB is privatized in 1990. Against this background of a variety of designs and operational responsibilities, there is clearly a great deal of scope for advances in fuel and fuel component technology. It should be noted, however, that the nuclear energy policy within the United Kingdom, particularly with regard to PWR plants, has been to adopt well proven designs wherever possible. Emphasis is therefore directed towards achieving the successful operation of conservative systems, with research and development work on advanced options for future implementation. The following sections give an overview of the areas where advanced designs are either in production or under development for each of the UK reactor systems in turn, together with an indication of possible future developments

  17. Options for treatment of legacy and advanced nuclear fuels

    OpenAIRE

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  18. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  19. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  20. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  1. Advanced Coal-Fueled Gas Turbine Program

    Energy Technology Data Exchange (ETDEWEB)

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  2. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  3. IEA-Advanced Motor Fuels Annual Report 2010

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-12-02

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes the agreement, activities, and projects for the year. A section on the global situation for Advanced Motor Fuels includes country reports from each participating AMF member. A status report on each active annex for the agreement is also included, as is a message from the AMF Chairman. Final sections include an Outlook for Advanced Motor Fuels, further information, and a glossary of terms.

  4. Advanced fuel system technology for utilizing broadened property aircraft fuels

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  5. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  6. Advanced methods of solid oxide fuel cell modeling

    CERN Document Server

    Milewski, Jaroslaw; Santarelli, Massimo; Leone, Pierluigi

    2011-01-01

    Fuel cells are widely regarded as the future of the power and transportation industries. Intensive research in this area now requires new methods of fuel cell operation modeling and cell design. Typical mathematical models are based on the physical process description of fuel cells and require a detailed knowledge of the microscopic properties that govern both chemical and electrochemical reactions. ""Advanced Methods of Solid Oxide Fuel Cell Modeling"" proposes the alternative methodology of generalized artificial neural networks (ANN) solid oxide fuel cell (SOFC) modeling. ""Advanced Methods

  7. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as high performance fuel concepts now under development for delivery in the late 1980s. The paper also discusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability

  8. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  9. Advanced Combustion and Fuels; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Zigler, Brad

    2015-06-08

    Presented at the U.S. Department of Energy Vehicle Technologies Office 2015 Annual Merit Review and Peer Evaluation Meeting, held June 8-12, 2015, in Arlington, Virginia. It addresses technical barriers of inadequate data and predictive tools for fuel and lubricant effects on advanced combustion engines, with the strategy being through collaboration, develop techniques, tools, and data to quantify critical fuel physico-chemical effects to enable development of advanced combustion engines that use alternative fuels.

  10. Epithermal interrogation of fissile waste

    Energy Technology Data Exchange (ETDEWEB)

    Coop, K.L.; Hollas, C.L.

    1996-09-01

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described.

  11. Epithermal interrogation of fissile waste

    International Nuclear Information System (INIS)

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

  12. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  13. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  14. TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    The separation of trivalent transplutonium actinides from fission product lanthanide ions represents a challenging aspect of advanced nuclear fuel partitioning schemes. The challenge of this separation could be amplified in the context of the AFCI-UREX+1a process, as Np and Pu will accompany the minor actinides to this stage of separation. At present, the baseline lanthanide-actinide separation method is the TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous complexes) process. TALSPEAK was developed in the late 1960's at Oak Ridge National Laboratory and has been demonstrated at pilot scale. This process relies on the complex interaction between an organic and an aqueous phase both containing complexants for selectively separating the trivalent actinide. The 3 complexing components are: the di(2-ethylhexyl) phosphoric acid (HDEHP), the lactic acid (HL) and the diethylenetriamine-N,N,N',N'',N''-pentaacetic acid (DTPA). In this report we discuss observations on kinetic and thermodynamic features described in the prior literature and describe some results of our ongoing research on basic chemical features of this system. The information presented indicates that the lactic acid buffer participates in the net operation of the TALSPEAK process in a manner that is not explained by existing information on the thermodynamic features if the known Eu(III)-lactate species. (authors)

  15. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  16. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  17. Status of advanced carbide fuels: Past, present, and future

    Science.gov (United States)

    Anghaie, Samim; Knight, Travis

    2002-01-01

    Solid solution, mixed uranium/refractory metal carbide fuels such as (U, Zr, Nb)C, so called ternary carbide or tri-carbide fuels have great potential for applications in next generation advanced nuclear power reactors. Because of their high melting points, high thermal conductivity, improved resistance to hot hydrogen corrosion, and good fission product retention, these advanced nuclear fuels have great potential for high performance reactors with increased safety margins. Despite these many benefits, some concerns regarding carbide fuels include compatibility issues with coolant and/or cladding materials and their endurance under the extreme conditions associated with nuclear thermal propulsion. The status of these fuels is reviewed to characterize their performance for space nuclear power applications. Results of current investigations are presented and as well as future directions of study for these advanced nuclear fuels. .

  18. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  19. Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

    Science.gov (United States)

    Burke, Kenneth A.

    2009-01-01

    The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

  20. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  1. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  2. Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A.; Tolk, K.

    2007-12-15

    This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.

  3. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  4. Non-proliferation, safeguards, and security for the fissile materials disposition program immobilization alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Duggan, R.A.; Jaeger, C.D.; Tolk, K.M. [Sandia National Labs., Albuquerque, NM (United States); Moore, L.R. [Lawrence Livermore National Lab., CA (United States)

    1996-05-01

    The Department of Energy is analyzing long-term storage and disposition alternatives for surplus weapons-usable fissile materials. A number of different disposition alternatives are being considered. These include facilities for storage, conversion and stabilization of fissile materials, immobilization in glass or ceramic material, fabrication of fissile material into mixed oxide (MOX) fuel for reactors, use of reactor based technologies to convert material into spent fuel, and disposal of fissile material using geologic alternatives. This paper will focus on how the objectives of reducing security and proliferation risks are being considered, and the possible facility impacts. Some of the areas discussed in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threats, and (4) potential proliferation, safeguards, and security issues and impacts on the facilities. Issues applicable to all of the possible disposition alternatives will be discussed in this paper. However, particular attention is given to the plutonium immobilization alternatives.

  5. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  6. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  7. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  8. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) is analyzing long-term storage on options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to bum the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the nonproliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  9. Criticality Control Fissile of Materials. Proceedings of the Symposium on Criticality Control of Fissile Materials

    International Nuclear Information System (INIS)

    Criticality control comprises all the administrative and technical procedures which enable the storage and processing of fissile material to be carried out under conditions of nuclear safety. It is of particular importance in the safe design and operation of chemical and metallurgical plants processing fissile material, in the handling and storage of enriched fuel for reactors, and in transportation of fissile material. The growth of nuclear power, with its increasing use of fissile material and production of plutonium, is leading to an ever widening need for this discipline. This Symposium was held 4½ years after the only other international meeting on this topic, at which the first broad exchange of ideas and theories enabled a comparison to be drawn between the various ways in which the subject is handled in the different countries. That meeting showed that criticality safety was often achieved by procedures known to be ultra-safe, as there was a great lack of useful experimental data with which to check theoretical models. Since that time the quantities of material being processed have increased, and with the now urgent necessity of achieving economic, and hence commercially competitive, operation, the procedure of using arbitrary factors of safety is no longer adequate. Plant Managers now require good data on the basis of which they can choose a suitable factor of safety, and design a process to be safe under any foreseeable circumstances. The present Symposium showed the great increase in the amount of available experimental data and its importance in checking the now highly sophisticated computer calculations. There are many diagrams in these Proceedings with curves from which critical parameters for various configurations can be taken. The dearth of data for plutonium systems is causing some difficulty in plutonium processing plants, which are becoming commercially important. The excellent safety record of the atomic energy industry was highlighted, and a

  10. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  11. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  12. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  13. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  14. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

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

  15. PNC`s proposal on the Advanced Fuel Recycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Kamiya, Masayoshi; Shinoda, Yoshihiko; Ojima, Hisao [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-03-01

    MOX fuel for FBR is allowed to contain impurities within several thousand ppm, which means less than 1000 of decontamination factor (DF) in reprocessing is enough for Pu and U recycle use. The Advanced Fuel Recycle proposed by PNC is on this basis. The concept consists of innovations on both MOX fuel fabrication and aqueous reprocessing technologies based on the Purex process and it is believed that successful optimization of fuel cycle interface condition is the key issue to realize the concept. The lower DF such as 1000 can be easily obtained by the simplified Purex flowsheet which has no purification steps. However, new subject arises in MOX fuel fabrication, that is, fabrication is conducted in the shielding cell using equipment which is maintained remotely. A simplified fabrication technology becomes essential to establish the remote maintenance system and is one of the critical path for achieving the Advanced Fuel Recycle. The PNC`s proposal on the advanced fuel recycle concept consists of modified PUREX process having single extraction cycle and crystallization, Remote fuel fabrication such as gelation and vibro-packing. In the Advanced Fuel Recycle concept, as it is low DF cycle system, all processes should be installed in remote maintenance cells. Then both reprocessing and fabrication facility would be able to be integrated into a same building. Integrated fuel cycle plant has several merits. No transportation of nuclear material between reprocessing and fabrication enhances non-proriferation aspect in addition to the low-DF concept. Cost performance is also improved because of optimization and rationalization of auxiliary equipment, and so on. (author)

  16. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

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

  17. Surrogate Model Development for Fuels for Advanced Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Anand, Krishnasamy [University of Wisconsin, Madison; Ra, youngchul [University of Wisconsin, Madison; Reitz, Rolf [University of Wisconsin; Bunting, Bruce G [ORNL

    2011-01-01

    The fuels used in internal-combustion engines are complex mixtures of a multitude of different types of hydrocarbon species. Attempting numerical simulations of combustion of real fuels with all of the hydrocarbon species included is highly unrealistic. Thus, a surrogate model approach is generally adopted, which involves choosing a few representative hydrocarbon species whose overall behavior mimics the characteristics of the target fuel. The present study proposes surrogate models for the nine fuels for advanced combustion engines (FACE) that have been developed for studying low-emission, high-efficiency advanced diesel engine concepts. The surrogate compositions for the fuels are arrived at by simulating their distillation profiles to within a maximum absolute error of 4% using a discrete multi-component (DMC) fuel model that has been incorporated in the multi-dimensional computational fluid dynamics (CFD) code, KIVA-ERC-CHEMKIN. The simulated surrogate compositions cover the range and measured concentrations of the various hydrocarbon classes present in the fuels. The fidelity of the surrogate fuel models is judged on the basis of matching their specific gravity, lower heating value, hydrogen/carbon (H/C) ratio, cetane number, and cetane index with the measured data for all nine FACE fuels.

  18. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  19. Fuel cell and advanced turbine power cycle

    Energy Technology Data Exchange (ETDEWEB)

    White, D.J. [Solar Turbines, Inc., San Diego, CA (United States)

    1995-10-19

    Solar Turbines, Incorporated (Solar) has a vested interest in the integration of gas turbines and high temperature fuel cells and in particular, solid oxide fuel cells (SOFCs). Solar has identified a parallel path approach to the technology developments needed for future products. The primary approach is to move away from the simple cycle industrial machines of the past and develop as a first step more efficient recuperated engines. This move was prompted by the recognition that the simple cycle machines were rapidly approaching their efficiency limits. Improving the efficiency of simple cycle machines is and will become increasingly more costly. Each efficiency increment will be progressively more costly than the previous step.

  20. Fissile solution dynamics: Student research

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.

    1994-09-01

    There are two research projects in criticality safety at the University of Arizona: one in dynamic simulation of hypothetical criticality accidents in fissile solutions, and one in criticality benchmarks using transport theory. We have used the data from nuclear excursions in KEWB, CRAC, and SILENE to help in building models for solution excursions. An equation of state for liquids containing gas bubbles has been developed and coupled to point-reactor dynamics in an attempt to predict fission rate, yield, pressure, and kinetic energy. It appears that radiolytic gas is unimportant until after the first peak, but that it does strongly affect the shape of the subsequent power decrease and also the dynamic pressure.

  1. ULTRACLEAN FUELS PRODUCTION AND UTILIZATION FOR THE TWENTY-FIRST CENTURY: ADVANCES TOWARDS SUSTAINABLE TRANSPORTATION FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, E.

    2013-06-17

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  2. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  3. Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

    International Nuclear Information System (INIS)

    R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

  4. Characterisation of fuels for advanced pressurised combustion

    Energy Technology Data Exchange (ETDEWEB)

    Zevenhoven, R.; Hupa, M.; Backman, P.; Forssen, M.; Karlsson, M.; Kullberg, M.; Sorvari, V.; Uusikartano, T. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group; Nurk, M. [Tallinskij Politekhnicheskij Inst., Tallinn (Estonia)

    1997-10-01

    The objective of the research was to determine a set of fuel characteristics which quantify the behaviour of a fuel in a typical pressurised combustor or gasifier environment, especially in hybrid processes such as second generation PFBC. One specific aspect was to cover a wide range of fuels, including several coal types and several grades of peat and biomasses: 7 types of coal, 2 types of peat, 2 types of wood, 2 types of black liquor, Estonian oil shale and Venezuelan Orimulsion were studied. The laboratory facilities used are a pressurised thermogravimetric reactor (PTGR), a pressurised grid heater (PGH) and an atmospheric entrained flow quartz tube reactor, with gas analysis, which can be operated as a fixed bed reactor. A major part of the work was related to fuel devolatilisation in the PGH and sequential devolatilisation and char gasification (with carbon dioxide or steam) in the PTGR. The final part of that work is reported here, with the combustion of Estonian oil shale at AFBC or PFBC conditions as additional subject. Devolatilisation of the fuels at atmospheric pressure in nitrogen while monitoring gaseous exhausts, followed by ultimate analysis of the chars has been reported earlier. Here, results on the analysis of the reduction of NO (with and without CO) on chars at atmospheric pressure in a fixed bed reactor are reported. Finally, a comparison is given between experimental results and direct numerical simulation with several computer codes, i.e. PyroSim, developed at TU Graz, Austria, and the codes Partikkeli, Pisara and Cogas, which were provided by VTT Energy, Jyvaeskylae

  5. Advanced Fuels Can Reduce the Cost of Getting Into Space

    Science.gov (United States)

    Palaszewski, Bryan A.

    1998-01-01

    Rocket propellant and propulsion technology improvements can reduce the development time and operational costs of new space vehicle programs, and advanced propellant technologies can make space vehicles safer and easier to operate, and can improve their performance. Five major areas have been identified for fruitful research: monopropellants, alternative hydrocarbons, gelled hydrogen, metallized gelled propellants, and high-energy-density propellants. During the development of the NASA Advanced Space Transportation Plan, these technologies were identified as those most likely to be effective for new NASA vehicles. Several NASA research programs had fostered work in fuels under the topic Fuels and Space Propellants for Reusable Launch Vehicles in 1996 to 1997. One component of this topic was to promote the development and commercialization of monopropellant rocket fuels, hypersonic fuels, and high-energy-density propellants. This research resulted in the teaming of small business with large industries, universities, and Government laboratories. This work is ongoing with seven contractors. The commercial products from these contracts will bolster advanced propellant research. Work also is continuing under other programs, which were recently realigned under the "Three Pillars" of NASA: Global Civil Aviation, Revolutionary Technology Leaps, and Access to Space. One of the five areas is described below, and its applications and effect on future missions is discussed. This work is being conducted at the NASA Lewis Research Center with the assistance of the NASA Marshall Space Flight Center. The regenerative cooling of spacecraft engines and other components can improve overall vehicle performance. Endothermic fuels can absorb energy from an engine nozzle and chamber and help to vaporize high-density fuel before it enters the combustion chamber. For supersonic and hypersonic aircraft, endothermic fuels can absorb the high heat fluxes created on the wing leading edges and

  6. LWR spent fuel storage technology: Advances and experience

    International Nuclear Information System (INIS)

    By 2003, the year the US Department of Energy (DOE) currently predicts a repository will be available, 58 domestic commercial nuclear-power plant units are expected to run out of wet storage space for LWR spent fuel. To alleviate this problem, utilities implemented advances in storage methods that increased storage capacity as well as reduced the rate of generating spent fuel. Those advances include (1) transhipping spent-fuel assemblies between pools within the same utility system, (2) reracking pools to accommodate additional spent-fuel assemblies, (3) taking credit for fuel burnup in pool storage rack designs, (4) extending fuel burnup, (5) rod consolidation, and (6) dry storage. The focus of this paper is on advances in rod consolidation and dry storage. Wet storage continues to be the predominant US spent-fuel management technology, but as a measure to enhance at-reactor storage capacity, the Nuclear Waste Policy Act of 1982 authorized DOE to assist utilities with licensing at-reactor dry storage. Information exchanges with other nations, laboratory testing and modeling, and cask tests cooperatively funded by US utilities and DOE produced a strong technical basis to develop confidence that LWR spent fuel can be stored safely for several decades in both wet and dry modes. Licensed dry storage of spent fuel in an inert atmosphere was first achieved in the US in 1986. Studies are underway in several countries to determine acceptable conditions for storing LWR spent fuel in air. Rod-consolidation technology is being developed and demonstrated to enhance the capacity for both wet and dry storage. Large-scale commercial implementation is awaiting optimization of practical and economical mechanical systems. 22 refs., 1 fig

  7. Strategic research of advanced fuel cycle technologies in JNC

    Energy Technology Data Exchange (ETDEWEB)

    Kawata, T.; Fukushima, M.; Nomura, S. [Japan Nuclear Cycle Development Institute, Tokai Works (Japan)

    2000-07-01

    Key technologies for the future nuclear fuel cycle have been proposed and are being reviewed in JNC as a part of the Feasibility Study for an Advanced Fuel Cycle, which is to achieve a more flexible energy choice to satisfy a sustainable energy security and global environmental protection. The candidate reprocessing technologies are: 1) aqueous simplified PUREX process, 2) oxide or metallic electrowinning, and 3) fluoride volatilization for oxide, metal, or nitride fuels. The fuel fabrication methods being investigated are: 1) simplified pellet process, 2) sphere/vibro-packed process for MOX/MN fuel, and 3) casting for metal fuel. These candidate technologies are currently being compared based on past experiences, technical issues to be solved, industrial applicability for future plants, feasible options for MA/LLFP separation, and nonproliferation aspects. Alter two years of the present reviewing process, selected key technologies will be developed over the next five years to evaluate industrial applicability of reprocessing and fuel manufacturing processes for the advanced fuel cycle. (authors)

  8. Strategic research of advanced fuel cycle technologies in JNC

    International Nuclear Information System (INIS)

    Key technologies for the future nuclear fuel cycle have been proposed and are being reviewed in JNC as a part of the Feasibility Study for an Advanced Fuel Cycle, which is to achieve a more flexible energy choice to satisfy a sustainable energy security and global environmental protection. The candidate reprocessing technologies are: 1) aqueous simplified PUREX process, 2) oxide or metallic electrowinning, and 3) fluoride volatilization for oxide, metal, or nitride fuels. The fuel fabrication methods being investigated are: 1) simplified pellet process, 2) sphere/vibro-packed process for MOX/MN fuel, and 3) casting for metal fuel. These candidate technologies are currently being compared based on past experiences, technical issues to be solved, industrial applicability for future plants, feasible options for MA/LLFP separation, and nonproliferation aspects. Alter two years of the present reviewing process, selected key technologies will be developed over the next five years to evaluate industrial applicability of reprocessing and fuel manufacturing processes for the advanced fuel cycle. (authors)

  9. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  10. Natural Gas for Advanced Dual-Fuel Combustion Strategies

    Science.gov (United States)

    Walker, Nicholas Ryan

    Natural gas fuels represent the next evolution of low-carbon energy feedstocks powering human activity worldwide. The internal combustion engine, the energy conversion device widely used by society for more than one century, is capable of utilizing advanced combustion strategies in pursuit of ultra-high efficiency and ultra-low emissions. Yet many emerging advanced combustion strategies depend upon traditional petroleum-based fuels for their operation. In this research the use of natural gas, namely methane, is applied to both conventional and advanced dual-fuel combustion strategies. In the first part of this work both computational and experimental studies are undertaken to examine the viability of utilizing methane as the premixed low reactivity fuel in reactivity controlled compression ignition, a leading advanced dual-fuel combustion strategy. As a result, methane is shown to be capable of significantly extending the load limits for dual-fuel reactivity controlled compression ignition in both light- and heavy-duty engines. In the second part of this work heavy-duty single-cylinder engine experiments are performed to research the performance of both conventional dual-fuel (diesel pilot ignition) and advanced dual-fuel (reactivity controlled compression ignition) combustion strategies using methane as the premixed low reactivity fuel. Both strategies are strongly influenced by equivalence ratio; diesel pilot ignition offers best performance at higher equivalence ratios and higher premixed methane ratios, whereas reactivity controlled compression ignition offers superior performance at lower equivalence ratios and lower premixed methane ratios. In the third part of this work experiments are performed in order to determine the dominant mode of heat release for both dual-fuel combustion strategies. By studying the dual-fuel homogeneous charge compression ignition and single-fuel spark ignition, strategies representative of autoignition and flame propagation

  11. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code

    International Nuclear Information System (INIS)

    Highlights: ► New, three-dimensional, parallel, multi-physics code to simulate fuel behavior in nominal operation. ► Fully-coupled thermomechanics for nominal operation and operation during transients. ► Isotopic depletion using Scale/ORIGEN-S within a fuel performance code. ► Leveraging of existing, validated material models from existing fuel performance codes. ► Initial validation evaluation of an advanced modeling and simulation code for fuel performance. - Abstract: The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code is a new, three-dimensional, multi-physics tool that uses state-of-the-art solution methods and validated nuclear fuel models to simulate the nominal operation and anticipated operational transients of nuclear fuel. The AMP Nuclear Fuel Performance code leverages existing validated material models from traditional fuel performance codes and the Scale/ORIGEN-S spent-fuel characterization code to provide an initial capability that is shown to be sufficiently accurate for a single benchmark problem and anticipated to be accurate for a broad range of problems. The thermomechanics foundation can be solved in a time-dependent or quasi-static approach with any variation of operator-split or fully-coupled solutions at each time step through interoperable interfaces to leading computational mathematics tools, including PETSc, Trilinos, and SUNDIALS. A baseline validation of the AMP Nuclear Fuel Performance code has been performed through the modeling of an experiment in the Halden Reactor Project (IFA-432) that demonstrates the integrated capability and provides a baseline of the initial accuracy of the software.

  12. Characterisation of fuels for advanced pressurized combustion

    Energy Technology Data Exchange (ETDEWEB)

    Zevenhoven, R.; Hupa, M.; Backman, P.; Karlsson, M.; Kullberg, M.; Sorvari, V. [Aabo Akademi, Turku (Finland); Nurk, M. [Tallinn Univ. (Estonia)

    1996-12-01

    After 2 of the 3 years for this EU Joule 2 extension project, a rough comparison on the devolatilisation behaviour and char reactivity of 11 fossil fuels and 4 biofuels has been obtained. The experimental plan for 1995 has been completed, the laboratory facilities appeared to be well suited for the broad range of analyses presented here. A vast amount of devolatilisation tests in nitrogen at atmospheric pressure with gas analysis and char analysis gave a lot of information on the release of carbon, sulphur, nitrogen and also sodium, chloride and some other elements. Also first-order rate parameters could be determined. Solid pyrolysis yield measurements with the pressurised grid heater show a very good reproducibility except for the fuels with high carbonate content and those with very small char yield. Problems have to be solved considering lower heating rates and the use of folded grids. Fuel pyrolysis followed by gasification (with carbon dioxide or water as oxidising agent) at various temperatures and pressures shows that in general char solid yields and gasification reactivities are higher at elevated pressure. The design and construction of a pressurized single particle reactor, to be operational early 1996 is currently being negotiated. Numerical modelling of coal devolatilisation shows that even for atmospheric pressures the results differ significantly from experimental findings. (author)

  13. Alternative Fuel and Advanced Technology Commercial Lawn Equipment

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-10-10

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  14. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-10-01

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  15. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  16. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  17. Design study on advanced reprocessing systems for FR fuel cycle

    International Nuclear Information System (INIS)

    A design study has been carried out for four advanced reprocessing technologies for the future fast rector (FR) recycle systems (advanced aqueous, and three non-aqueous systems based on oxide electrowinning, metal electrorefining, and fluoride volatility methods). The systems were evaluated mainly from the viewpoint of economics. It has been shown that, for MOX fuel reprocessing, all the systems with a capacity of 200 t/y attains the economical target, whereas for such a small capacity as 50 t/y, only the non-aqueous systems have potential to attain the target. For metallic and nitride fuel, a metal electrorefining system has been shown to be advantageous. (author)

  18. IEA-Advanced Motor Fuels Annual Report 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-06-15

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  19. IEA-Advanced Motor Fuels Annual Report 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  20. Advanced PWR fuel assembly development programs in Korea

    International Nuclear Information System (INIS)

    Both KNFC and Westinghouse have continued to focus on developing products that will meet the challenge of increasing fuel duty requirements in Korea. These higher duty conditions include higher energy core designs through improved plant capacity factors, power uprate, extended fuel burnup, peaking factor increases, and more severe coolant chemistry (including high lithium concentration). Recent advanced fuel development activities in Korea include implementation of the 17x17 Robust Fuel Assembly (RFA), which is currently in operation with excellent performance in the United States and Europe, as well as the 16x16 PLUS7TM fuel assembly for use in KSNP plants. KNFC and Westinghouse are jointly developing advanced fuel that will meet future fuel duty challenges of 17x17 and 16x16 Westinghouse type plants. This paper focuses on advanced fuel assembly development programs that are underway and how these designs demonstrate improved margins under high duty plant operating conditions. In designing for these high duty conditions key design considerations for the various operational modes (i.e. power uprating, high burnup, long cycles, etc.) must be identified. These design considerations will include the traditional factors such as safety margin (DNB and LOCA), fuel rod design margin (e.g. corrosion, internal pressure, etc.) and mechanical design margins, among others. In addressing these design considerations, the fundamental approach is to provide additional design margin through materials, mechanical, and thermal performance enhancements, to assure flawless fuel performance. The foundation of all fuel designs is the product development process used to meet the demands of modern high duty operation including power uprating, high burnup, longer cycles, and high-lithium coolant chemistries. These advanced fuel assembly designs incorporate features that provide improved mechanical design margin, as well as thermal performance margin (DNB). Enhanced grid designs result in a

  1. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  2. Structural evaluation of Siemens advanced fuel channel under accident loadings

    International Nuclear Information System (INIS)

    As a part of an effort to develop an advanced BWR fuel channel design, Siemens Power Corporation (SPC) and the Siemens AG Power Generation Group (KWU) performed structural analyses to verify the acceptability of the fuel channel design under combined seismic/LOCA (Loss Of. Coolant Accident) loadings. The results of the analyses give some interesting insights into the problem: 1) fluid-structure interaction (FSI) effects are significant and should be considered, 2) the problem may simplified by using a linear analysis despite non-linear features (gaps) between interfacing components, and 3) sufficient accuracy may be obtained by using only the first mode of vibration. The channeled fuel assembly can be considered to be a beam where the flexural stiffness is primarily determined by the fuel channel and the mass is given by the fuel assembly. The results from the analyses show the advanced fuel channel design meets applicable design criteria with adequate margins while at the same time exhibiting superior nuclear performance compared to a conventional BWR fuel channel. (author)

  3. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. the technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy environment. the world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, Candu reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuel which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the Candu reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential Candu fuel cycle developments can be accommodated in existing

  4. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO2 pellet-pins. The advanced PHWR fuels are UO2-PuO2 (≤ 2 per cent), ThO2-PuO2 (≤ 4 per cent) and ThO2-U233O2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO2, PuO2 and ThO2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  5. Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Wang, M. Q.

    1998-12-16

    At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions.

  6. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

  7. An analysis on the breeding capability and safety related parameters of advanced fast reactor fuels using recent cross-section set

    International Nuclear Information System (INIS)

    Highlights: • Breeding ratio of fast reactor fuels is computed with latest cross-section set. • Safety related parameters are also evaluated. • It is found that there are better prospects of utilization of thorium resources. • With large fast reactors, Th–233U fuel combination gives better B.G. -- Abstract: This study focuses on the evaluation of breeding capability as well as safety related neutronic parameters of advanced fast reactor fuels which comprises of fissile–fertile combination of metal, oxide, carbide and nitride, using the recent neutron cross-section set ENDF/B-VI.7. Sodium cooled fast breeder reactor similar to prototype Fast Breeder Reactor (PFBR) is used to evaluate the performance of various fuel types involving fissile isotopes of 233U and Pu and fertile isotopes of Th and 238U. The analysis is restricted to a comparison of neutronic parameters of a fresh core and does not take into account the effects of burnup and fission products. The breeding potential of the fuels are also compared with European cross-section set JEFF-3.1. The breeding ratio of advanced fuels evaluated with ENDF/B-VI.7 and JEFF-3.1 was found to be in good agreement. From this study, it is found that Th–233U combination for almost all fuel types with the present geometry and composition gives a lower breeding ratio value. Safety neutronic parameters such as effective delayed neutron fraction, Doppler defect and sodium void reactivity were also computed. In terms of breeding potential and safety neutronic parameters, the performance of Th–Pu system especially the metal fuel type can be a better option for future large fast reactors. The large negative Doppler feedback along with a negative sodium void reactivity for metal and hybrid combinations of Th–233U system makes it an attractive fuel cycle option even though there is a penalty over its breeding capability

  8. Fissile material containment efforts: an overview

    International Nuclear Information System (INIS)

    Fissile materials (plutonium and highly enriched uranium (HEU)) are the fundamental ingredients of all nuclear weapons and they also happen to be the most difficult and expensive part of a nuclear warhead to produce. For a comprehensive nuclear disarmament and non-proliferation regime to be established, it becomes essential that there is a global, verified ban on the production of fissile materials for nuclear explosives

  9. Introducing advanced thorium-based fuel cycles in SA : an evolutionary approach / Maria Hendrina (Marina) du Toit

    OpenAIRE

    Du Toit, Maria Hendrina

    2013-01-01

    Past experience in several thorium fuelled research- and power reactors provides the basis and history of thorium. The material properties, fertile- and fissile isotope properties as well as the decay chain of thorium are discussed for purposes of evaluating thorium as a source of fuel. The different thorium-based fuel designs for PWR cores are discussed and resulting difficulties and solutions are outlined. The different options for each strategy are compared in terms of ad...

  10. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO2, lower sensitivity to 234U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  11. Advanced materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Stevenson, J.

    1995-08-01

    The purpose of this research is to improve the properties of the current state-of-the-art materials used for solid oxide fuel cells (SOFCs). The objectives are to: (1) develop materials based on modifications of the state-of-the-art materials; (2) minimize or eliminate stability problems in the cathode, anode, and interconnect; (3) Electrochemically evaluate (in reproducible and controlled laboratory tests) the current state-of-the-art air electrode materials and cathode/electrolyte interfacial properties; (4) Develop accelerated electrochemical test methods to evaluate the performance of SOFCs under controlled and reproducible conditions; and (5) Develop and test materials for use in low-temperature SOFCs. The goal is to modify and improve the current state-of-the-art materials and minimize the total number of cations in each material to avoid negative effects on the materials properties. Materials to reduce potential deleterious interactions, (3) improve thermal, electrical, and electrochemical properties, (4) develop methods to synthesize both state-of-the-art and alternative materials for the simultaneous fabricatoin and consolidation in air of the interconnections and electrodes with the solid electrolyte, and (5) understand electrochemical reactions at materials interfaces and the effects of component composition and processing on those reactions.

  12. Assessment of Research Needs for Advanced Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Penner, S.S.

    1985-11-01

    The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

  13. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  14. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  15. Design study and evaluation of advanced fuel fabrication systems for FBR fuel cycle

    International Nuclear Information System (INIS)

    The conceptual design study for advanced FBR fuel fabrication system has been performed for the purpose that the feature of small-scale fabrication system in the transition stage from LWR to FBR fuel cycle. On the small-scale of 50 ton heavy metal per year fabrication system, dry type fabrication systems have superior cost performance than the wet type, although waste amount is larger. (authors)

  16. Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

    International Nuclear Information System (INIS)

    The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO2. The model was initially tested and the average discharge burnup for natural UO2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

  17. Development of Advanced Voloxidation Process for Treatment of Spent Fuel

    International Nuclear Information System (INIS)

    Data for evaluation of the effects of advanced voloxidation on pyroprocessing of spent oxide fuel with a determination for a path forward such was produced as follows: effect of particle size and particle structure on oxide reduction, assessment of decladding options for pyroprocessing, effect of removal timing of fission products, analysis of radioactivity and decay heat of advanced voloxidation process, proliferation resistance of advanced voloxidation process, Effect of advanced voloxidation process on shielding. Also, performance objectives for advanced voloxidation with respective to the down stream effects was established. The technology on design and manufacture of voloxidation and off gas treatment equipment was established. The possibility of fabrication of porous granule as a feed material for electro-reduction process was confirmed using rotary voloxidizer and SIMFUEL. The operational conditions for advanced voloxidation process consisting of 4 steps heat treatment was drawn to vaporize fission products and fabricate UO2 granule. The trapping test of Cs and Re(surrogate material of Tc) using newly developed filter were selectively separated at trapping efficiency of 99%, respectively. Data for oxidative decladding, vaporization rate of fission products, and particle size from experiment on voloxidation using spent fuel in ILN hot cell was acquisited including data of off gas trapping characteristics and verification of excellent performance of filter

  18. Cycle update : advanced fuels and technologies for emissions reduction

    Energy Technology Data Exchange (ETDEWEB)

    Smallwood, G. [National Research Council of Canada, Ottawa, ON (Canada)

    2009-07-01

    This paper provided a summary of key achievements of the Program of Energy Research and Development advanced fuels and technologies for emissions reduction (AFTER) program over the funding cycle from fiscal year 2005/2006 to 2008/2009. The purpose of the paper was to inform interested parties of recent advances in knowledge and in science and technology capacities in a concise manner. The paper discussed the high level research and development themes of the AFTER program through the following 4 overarching questions: how could advanced fuels and internal combustion engine designs influence emissions; how could emissions be reduced through the use of engine hardware including aftertreatment devices; how do real-world duty cycles and advanced technology vehicles operating on Canadian fuels compare with existing technologies, models and estimates; and what are the health risks associated with transportation-related emissions. It was concluded that the main issues regarding the use of biodiesel blends in current technology diesel engines are the lack of consistency in product quality; shorter shelf life of biodiesel due to poorer oxidative stability; and a need to develop characterization methods for the final oxygenated product because most standard methods are developed for hydrocarbons and are therefore inadequate. 2 tabs., 13 figs.

  19. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h)-1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  20. Construction and engineering report for advanced nuclear fuel development facility

    International Nuclear Information System (INIS)

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m2, basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

  1. Construction and engineering report for advanced nuclear fuel development facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m{sup 2}, basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc.

  2. A worldwide program for controlling fissile material

    International Nuclear Information System (INIS)

    Fissile (fissionable) material, either plutonium or highly enriched uranium or both in combination, is an essential ingredient of any nuclear weapon. This is equally true of the first generation of pure fission weapons - such as those used on Hiroshima and Nagasaki - and more modern weapons which gain part of their yield from thermonuclear fusion. In the latter case, a fission explosion is required to ignite the fusion reaction. This paper proposes a worldwide programme based upon that linkage. Within this programme, however, the dominant element would be a cut-off in the production of fissile material for nuclear weapons. This is a subject to which various analysts have devoted attention over the years. Most notably, recent work by analysts at Princeton University in the USA has convincingly shown that an initial cut-off of production in the USA and the Soviet Union could be adequately verified. In recognition of this work, the approach taken here is to build upon the Princeton analysis, showing how the verification regime appropriate to an initial US-Soviet cut-off could be expanded to cover reductions in military fissile materials inventories in those two countries, production cut-offs and inventory reductions in other NWSs, and worldwide regulation of civilian programmes for fissile material production. The description of this ultimate fissile material control regime is couched primarily in terms of the criteria according to which the regime should be designed, and the nature of the institutions and agreements which should support it

  3. Advanced modeling of oxy-fuel combustion of natural gas

    Energy Technology Data Exchange (ETDEWEB)

    Chungen Yin

    2011-01-15

    The main goal of this small-scale project is to investigate oxy-combustion of natural gas (NG) through advanced modeling, in which radiation, chemistry and mixing will be reasonably resolved. 1) A state-of-the-art review was given regarding the latest R and D achievements and status of oxy-fuel technology. The modeling and simulation status and achievements in the field of oxy-fuel combustion were also summarized; 2) A computer code in standard c++, using the exponential wide band model (EWBM) to evaluate the emissivity and absorptivity of any gas mixture at any condition, was developed and validated in detail against data in literature. A new, complete, and accurate WSGGM, applicable to both air-fuel and oxy-fuel combustion modeling and applicable to both gray and non-gray calculation, was successfully derived, by using the validated EWBM code as the reference mode. The new WSGGM was implemented in CFD modeling of two different oxy-fuel furnaces, through which its great, unique advantages over the currently most widely used WSGGM were demonstrated. 3) Chemical equilibrium calculations were performed for oxy-NG flame and air-NG flame, in which dissociation effects were considered to different degrees. Remarkable differences in oxy-fuel and air-fuel combustion were revealed, and main intermediate species that play key roles in oxy-fuel flames were identified. Different combustion mechanisms are compared, e.g., the most widely used 2-step global mechanism, refined 4-step global mechanism, a global mechanism developed for oxy-fuel using detailed chemical kinetic modeling (CHEMKIN) as reference. 4) Over 15 CFD simulations were done for oxy-NG combustion, in which radiation, chemistry, mixing, turbulence-chemistry interactions, and so on were thoroughly investigated. Among all the simulations, RANS combined with 2-step and refined 4-step mechanism, RANS combined with CHEMKIN-based new global mechanism for oxy-fuel modeling, and LES combined with different combustion

  4. The economics of advanced fuel cycles in CANDU (PHW) reactors

    International Nuclear Information System (INIS)

    The economic assessments of advanced fuel cycles performed within Ontario Hydro are collated and summarized. The results of the analyses are presented in a manner designed to provide a broad perspective of the economic issues regarding the advanced cycles. The enriched uranium fuel cycle is shown to be close to competitive at today's uranium prices, and its relative position vis-a-vis the natural uranium cycle will improve as uranium prices continue to rise. In the longer term, the plutonium-topped thorium cycle is identified as being the most economically desirable. It is suggested that this cycle may not be commercially attractive until the second or third decade of the next century. (auth)

  5. Advanced materials for alternative fuel capable directly fired heat engines

    Energy Technology Data Exchange (ETDEWEB)

    Fairbanks, J.W.; Stringer, J. (eds.)

    1979-12-01

    The first conference on advanced materials for alternative fuel capable directly fired heat engines was held at the Maine Maritime Academy, Castine, Maine. It was sponsored by the US Department of Energy, (Assistant Secretary for Fossil Energy) and the Electric Power Research Institute, (Division of Fossil Fuel and Advanced Systems). Forty-four papers from the proceedings have been entered into EDB and ERA and one also into EAPA; three had been entered previously from other sources. The papers are concerned with US DOE research programs in this area, coal gasification, coal liquefaction, gas turbines, fluidized-bed combustion and the materials used in these processes or equipments. The materials papers involve alloys, ceramics, coatings, cladding, etc., and the fabrication and materials listing of such materials and studies involving corrosion, erosion, deposition, etc. (LTN)

  6. The US Advanced Fuel Cycle Programme: Objectives and Accomplishments

    International Nuclear Information System (INIS)

    For approximately a decade, the United States Department of Energy has been conducting an advanced fuel cycle programme, presently named the Fuel Cycle R and D Program, devoted to lessening both the environmental burden of nuclear energy and the proliferation risk of accumulating used nuclear fuel. Currently, the programme is being redirected towards a science based, goal oriented focus with the objective of deploying successfully demonstrated technology in the 2040-2050 time frame. The present paper reports the key considerations of the science based research approach, the elements of the technical programme and the accomplishments in fast reactor research and development, the goal of which is to improve the primary issues that have inhibited fast reactor introduction in the past, namely, economics and safety. (author)

  7. Advanced coal-fueled industrial cogeneration gas turbine system

    Energy Technology Data Exchange (ETDEWEB)

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; Wen, C.S.

    1991-07-01

    Advances in coal-fueled gas turbine technology over the past few years, together with recent DOE-METC sponsored studies, have served to provide new optimism that the problems demonstrated in the past can be economically resolved and that the coal-fueled gas turbine can ultimately be the preferred system in appropriate market application sectors. The objective of the Solar/METC program is to prove the technical, economic, and environmental feasibility of a coal-fired gas turbine for cogeneration applications through tests of a Centaur Type H engine system operated on coal fuel throughout the engine design operating range. The five-year program consists of three phases, namely: (1) system description; (2) component development; (3) prototype system verification. A successful conclusion to the program will initiate a continuation of the commercialization plan through extended field demonstration runs.

  8. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  9. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    International Nuclear Information System (INIS)

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO2, PuO2 or (U,Pu)O2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO2 and UO2-PuO2) or of the precipitating crystalIization (PuO2). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O2, (U,Pu,Np)O2, (U,Am)O2 and (U,Pu,Am)O2. Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  10. Resonance self-indication as a method for estimating fissile elements in spent nuclear assemblies

    International Nuclear Information System (INIS)

    The integral version of the neutron resonance transmission analysis (NRTA) has been developed in order to solve the problem of estimating fissile elements in spent nuclear fuel. Two variants of the integral neutron resonance transmission analysis (the resonance self-indication (the average transmission by fission) and the average full transmission) have been described

  11. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  12. Systematic analysis of advanced fusion fuel in inertial fusion energy

    Science.gov (United States)

    Velarde, G.; Eliezer, S.; Henis, Z.; Piera, M.; Martinez-Val, J. M.

    1997-04-01

    Aneutronic fusion reactions can be considered as the cleanest way to exploit nuclear energy. However, these reactions present in general two main drawbacks.—very high temperatures are needed to reach relevant values of their cross sections—Moderate (and even low) energy yield per reaction. This value is still lower if measured in relation to the Z number of the reacting particles. It is already known that bremsstrahlung overruns the plasma reheating by fusion born charged-particles in most of the advanced fuels. This is for instance the case for proton-boron-11 fusion in a stoichiometric plasma and is also so in lithium isotopes fusion reactions. In this paper, the use of deuterium-tritium seeding is suggested to allow to reach higher burnup fractions of advanced fuels, starting at a lower ignition temperature. Of course, neutron production increases as DT contents does. Nevertheless, the ratio of neutron production to energy generation is much lower in DT-advanced fuel mixtures than in pure DT plasmas. One of the main findings of this work is that some natural resources (as D and Li-7) can be burned-up in a catalytic regime for tritium. In this case, neither external tritium breeding nor tritium storage are needed, because the tritium inventory after the fusion burst is the same as before it. The fusion reactor can thus operate on a pure recycling of a small tritium inventory.

  13. Combustion behaviors of a compression-ignition engine fueled with diesel/methanol blends under various fuel delivery advance angles.

    Science.gov (United States)

    Huang, Zuohua; Lu, Hongbing; Jiang, Deming; Zeng, Ke; Liu, Bing; Zhang, Junqiang; Wang, Xibin

    2004-12-01

    A stabilized diesel/methanol blend was described and the basic combustion behaviors based on the cylinder pressure analysis was conducted in a compression-ignition engine. The study showed that increasing methanol mass fraction of the diesel/methanol blends would increase the heat release rate in the premixed burning phase and shorten the combustion duration of the diffusive burning phase. The ignition delay increased with the advancing of the fuel delivery advance angle for both the diesel fuel and the diesel/methanol blends. For a specific fuel delivery advance angle, the ignition delay increased with the increase of the methanol mass fraction (oxygen mass fraction) in the fuel blends and the behaviors were more obvious at low engine load and/or high engine speed. The rapid burn duration and the total combustion duration increased with the advancing of the fuel delivery advance angle. The centre of the heat release curve was close to the top-dead-centre with the advancing of the fuel delivery advance angle. Maximum cylinder gas pressure increased with the advancing of the fuel delivery advance angle, and the maximum cylinder gas pressure of the diesel/methanol blends gave a higher value than that of the diesel fuel. The maximum mean gas temperature remained almost unchanged or had a slight increase with the advancing of the fuel delivery advance angle, and it only slightly increased for the diesel/methanol blends compared to that of the diesel fuel. The maximum rate of pressure rise and the maximum rate of heat release increased with the advancing of the fuel delivery advance angle of the diesel/methanol blends and the value was highest for the diesel/methanol blends.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-01

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

  15. Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

    2013-09-03

    ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

  16. Studies of neutron methods for process control and criticality surveillance of fissile material processing facilities

    International Nuclear Information System (INIS)

    The development of radiochemical processes for fissile material processing and spent fuel handling need new control procedures enabling an improvement of plant throughput. This is strictly related to the implementation of continuous criticality control policy and developing reliable methods for monitoring the reactivity of radiochemical plant operations in presence of the process perturbations. Neutron methods seem to be applicable for fissile material control in some technological facilities. The measurement of epithermal neutron source multiplication with heuristic evaluation of measured data enables surveillance of anomalous reactivity enhancement leading to unsafe states. 80 refs., 47 figs., 33 tabs. (author)

  17. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  18. Ignition timing advance in the bi-fuel engine

    Directory of Open Access Journals (Sweden)

    Marek FLEKIEWICZ

    2009-01-01

    Full Text Available The influence of ignition timing on CNG combustion process has been presented in this paper. A 1.6 liter SI engine has been tested in the special program. For selected engine operating conditions, following data were acquired: in cylinder pressure, crank angle, fuel mass consumption and exhaust gases temperatures. For the timing advance correction varying between 0 to 15 deg crank angle, the internal temperature of combustion chamber, as well as the charge combustion ratio and ratio of heat release has been estimated. With the help of the mathematical model, emissions of NO, CO and CO2 were additionally estimated. Obtained results made it possible to compare the influence of ignition timing advance on natural gas combustion in the SI engine. The engine torque and in-cylinder pressure were used for determination of the optimum engine timing advance.

  19. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  20. Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

    2013-09-03

    ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

  1. Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

    KAUST Repository

    Burger, Jessica L.

    2015-07-16

    © This article not subject to U.S. Copyright. Published 2015 by the American Chemical Society. Incremental but fundamental changes are currently being made to fuel composition and combustion strategies to diversify energy feedstocks, decrease pollution, and increase engine efficiency. The increase in parameter space (by having many variables in play simultaneously) makes it difficult at best to propose strategic changes to engine and fuel design by use of conventional build-and-test methodology. To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council (CRC) and other entities, so that researchers in different laboratories have access to fuels with consistent properties. In this work, six gasolines (FACE A, C, F, G, I, and J) are characterized by the advanced distillation curve (ADC) method to determine the composition and enthalpy of combustion in various distillate volume fractions. Tracking the composition and enthalpy of distillate fractions provides valuable information for determining structure property relationships, and moreover, it provides the basis for the development of equations of state that can describe the thermodynamic properties of these complex mixtures and lead to development of surrogate fuels composed of major hydrocarbon classes found in target fuels.

  2. Development of Kinetic Mechanisms for Next-Generation Fuels and CFD Simulation of Advanced Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNenly, Matt J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, Russell [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Killingsworth, Nick J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-17

    Predictive chemical kinetic models are needed to represent next-generation fuel components and their mixtures with conventional gasoline and diesel fuels. These kinetic models will allow the prediction of the effect of alternative fuel blends in CFD simulations of advanced spark-ignition and compression-ignition engines. Enabled by kinetic models, CFD simulations can be used to optimize fuel formulations for advanced combustion engines so that maximum engine efficiency, fossil fuel displacement goals, and low pollutant emission goals can be achieved.

  3. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

  4. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  5. Nuclear archaeology: Verifying declarations of fissile-material production

    International Nuclear Information System (INIS)

    Controlling the production of fissile material is an essential element of nonproliferation policy. Similarly, accounting for the past production of fissile material should be an important component of nuclear disarmament. This paper describes two promising techniques that make use of physical evidence at reactors and enrichment facilities to verify the past production of plutonium and highly enriched uranium. In the first technique, the concentrations of long-lived radionuclides in permanent components of the reactor core are used to estimate the neutron fluence in various regions of the reactor, and thereby verify declarations of plutonium production in the reactor. In the second technique, the ratio of the concentration of U-235 to that of U-234 in the tails is used to determine whether a given container of tails was used in the production of low- enriched uranium, which is suitable for reactor fuel, or highly enriched uranium, which can be used in nuclear weapons. Both techniques belong to the new field of open-quotes nuclear archaeology,close quotes in which the authors attempt to document past nuclear weapons activities and thereby lay a firm foundation for verifiable nuclear disarmament. 11 refs., 1 fig., 3 tabs

  6. The trafficking control of fissile and radioactive materials in Kazakhstan

    International Nuclear Information System (INIS)

    After disintegration of the USSR the problem of control for transfers of radioactive and fissile materials has getting more and more urgent in sovereign Kazakhstan. Fissile materials, which theoretically could be used to creation of nuclear weapons, were available at the following enterprises of Kazakhstan: The Institute of Nuclear Physics - National Nuclear Center - Almaty The Institute of Atomic Energy - National Nuclear Center - Kurchatov Mangyshlak Atomic Energy Combine - Aktau Ulbinskiy Metallurgical Plant - Ust-Kamenogorsk At last time mass-media have reported many incidents of a new type of theft. Individuals are detained, trying to sell fuel pellets made of enriched uranium, radioactive contaminated cables, or various components of technological purposes from the enterprises. A significant problem is the legal and illegal collection of ferrous and non-ferrous metals. This made at places of realization of nuclear tests (former Semipalatinsk nuclear test-site (STS), Azgir, Lyra and others) and at working and closed uranium mines. Nominally, the incidents can be divided into three groups: - heft of materials, hazardous in term of nonproliferation; - heft of materials, non-hazardous in terms of nonproliferation (radioactive scrap metal, sealed radioactive sources); - heft due to an inefficient system of physical protection (the tank for transportation of radioactive waste products at Institute of Atomic Energy of NNC RK)

  7. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  8. Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

  9. High efficiency fuel cell/advanced turbine power cycles

    Energy Technology Data Exchange (ETDEWEB)

    Morehead, H. [Westinghouse Electric Corp., Orlando, FL (United States)

    1995-10-19

    An outline of the Westinghouse high-efficiency fuel cell/advanced turbine power cycle is presented. The following topics are discussed: The Westinghouse SOFC pilot manufacturing facility, cell scale-up plan, pressure effects on SOFC power and efficiency, sureCell versus conventional gas turbine plants, sureCell product line for distributed power applications, 20 MW pressurized-SOFC/gas turbine power plant, 10 MW SOFC/CT power plant, sureCell plant concept design requirements, and Westinghouse SOFC market entry.

  10. Advanced Diagnostics in Oxy-Fuel Combustion Processes

    DEFF Research Database (Denmark)

    Brix, Jacob; Toftegaard, Maja Bøg; Clausen, Sønnik;

    This report sums up the findings in PSO-project 010069, “Advanced Diagnostics in Oxy- Fuel Combustion Processes”. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory...... equipment. The use of the IR technique for determination of particle temperatures, particle sizes, and number density proved reliable in both the swirl burner and the laboratory scale fixed bed reactor. When the technique was used in the swirl burner the subsequent data treatment was sensitive to optical...

  11. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)

  12. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  13. THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

    Energy Technology Data Exchange (ETDEWEB)

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

  14. 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.

  15. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process Facility(ACPF) and hotcell system technologies were developed in this program for demonstrating safely and effectively the Advanced spent fuel Conditioning Process(ACP) on a lab-scale. With the analysis of work flow and characteristics of the process, ACP was successively demonstrated on a lab-scale experiments and the performance of process was evaluated. The hotcell system was comprehensively evaluated with those results and the design data for the engineering-scale demonstration was derived to propose the direction for the future research and development. The main items performed in this project were as follows. - The reconstruction of ACPF hotcell and demonstration of the ACP - The design and operation technologies for α-γ type nuclear hot cell facility - The overall evaluation of the performance, safety and operation ability of the hotcell system - The acquisition of the government licences for construction and operation and the IAEA licence for using nuclear materials The results of safety analysis and environmental effects assessment and performance data for ACPF had been used for acquiring the government licence for facility operation. The valuable experiences on pyroprocess facility design and operation knowledges would be applied to new Mock-up Facility being scheduled to be a previous stage facility of Integrated Pyroprocess Facility

  16. Advanced coal gasifier-fuel cell power plant systems design

    Science.gov (United States)

    Heller, M. E.

    1983-01-01

    Two advanced, high efficiency coal-fired power plants were designed, one utilizing a phosphoric acid fuel cell and one utilizing a molten carbonate fuel cell. Both incorporate a TRW Catalytic Hydrogen Process gasifier and regenerator. Both plants operate without an oxygen plant and without requiring water feed; they, instead, require makeup dolomite. Neither plant requires a shift converter; neither plant has heat exchangers operating above 1250 F. Both plants have attractive efficiencies and costs. While the molten carbonate version has a higher (52%) efficiency than the phosphoric acid version (48%), it also has a higher ($0.078/kWh versus $0.072/kWh) ten-year levelized cost of electricity. The phosphoric acid fuel cell power plant is probably feasible to build in the near term: questions about the TRW process need to be answered experimentally, such as weather it can operate on caking coals, and how effective the catalyzed carbon-dioxide acceptor will be at pilot scale, both in removing carbon dioxide and in removing sulfur from the gasifier.

  17. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  18. Advanced fuel cells for transportation applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-10

    This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.

  19. Advanced Coal-Fueled Gas Turbine Program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  20. AECL programs in advanced systems research

    International Nuclear Information System (INIS)

    The AECL program in advanced systems research is directed in the long term to securing the option of obtaining fissile fuel by electronuclear breeding (accelerator breeder or fusion breeder) and to providing a basis from which AECL might move into stand alone fusion energy if warranted. In the short term the program is directed to reaping benefits from electronuclear technology. This report outlines the main activities and research facilities in both the long-term and short-term subprograms

  1. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    International Nuclear Information System (INIS)

    Develops the tools to investigate the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  2. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    Energy Technology Data Exchange (ETDEWEB)

    Paul Wilson

    2009-11-02

    Develops the tools to investiage the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  3. State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems

    International Nuclear Information System (INIS)

    Development of innovative fuels such as homogeneous and heterogeneous fuels, ADS fuels, and oxide, metal, nitride and carbide fuels is an important stage in the implementation process of advanced nuclear systems. Several national and international R and D programmes are investigating minor actinide-bearing fuels due to their ability to help reduce the radiotoxicity of spent fuel and therefore decrease the burden on geological repositories. Minor actinides can be converted into a suitable fuel form for irradiation in reactor systems where they are transmuted into fission products with a significantly shorter half-life. This report compares recent studies of fuels containing minor actinides for use in advanced nuclear systems. The studies review different fuels for several types of advanced reactors by examining various technical issues associated with fabrication, characterisation, irradiation performance, design and safety criteria, as well as technical maturity. (authors)

  4. Advanced proton-exchange materials for energy efficient fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Cy H.; Grest, Gary Stephen; Hickner, Michael A.; Cornelius, Christopher James; Staiger, Chad Lynn; Hibbs, Michael R.

    2005-12-01

    The ''Advanced Proton-Exchange Materials for Energy Efficient Fuel Cells'' Laboratory Directed Research and Development (LDRD) project began in October 2002 and ended in September 2005. This LDRD was funded by the Energy Efficiency and Renewable Energy strategic business unit. The purpose of this LDRD was to initiate the fundamental research necessary for the development of a novel proton-exchange membranes (PEM) to overcome the material and performance limitations of the ''state of the art'' Nafion that is used in both hydrogen and methanol fuel cells. An atomistic modeling effort was added to this LDRD in order to establish a frame work between predicted morphology and observed PEM morphology in order to relate it to fuel cell performance. Significant progress was made in the area of PEM material design, development, and demonstration during this LDRD. A fundamental understanding involving the role of the structure of the PEM material as a function of sulfonic acid content, polymer topology, chemical composition, molecular weight, and electrode electrolyte ink development was demonstrated during this LDRD. PEM materials based upon random and block polyimides, polybenzimidazoles, and polyphenylenes were created and evaluated for improvements in proton conductivity, reduced swelling, reduced O{sub 2} and H{sub 2} permeability, and increased thermal stability. Results from this work reveal that the family of polyphenylenes potentially solves several technical challenges associated with obtaining a high temperature PEM membrane. Fuel cell relevant properties such as high proton conductivity (>120 mS/cm), good thermal stability, and mechanical robustness were demonstrated during this LDRD. This report summarizes the technical accomplishments and results of this LDRD.

  5. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  6. Advanced Fuel Cycle Economic Tools, Algorithms, and Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2009-05-01

    The Advanced Fuel Cycle Initiative (AFCI) Systems Analysis supports engineering economic analyses and trade-studies, and requires a requisite reference cost basis to support adequate analysis rigor. In this regard, the AFCI program has created a reference set of economic documentation. The documentation consists of the “Advanced Fuel Cycle (AFC) Cost Basis” report (Shropshire, et al. 2007), “AFCI Economic Analysis” report, and the “AFCI Economic Tools, Algorithms, and Methodologies Report.” Together, these documents provide the reference cost basis, cost modeling basis, and methodologies needed to support AFCI economic analysis. The application of the reference cost data in the cost and econometric systems analysis models will be supported by this report. These methodologies include: the energy/environment/economic evaluation of nuclear technology penetration in the energy market—domestic and internationally—and impacts on AFCI facility deployment, uranium resource modeling to inform the front-end fuel cycle costs, facility first-of-a-kind to nth-of-a-kind learning with application to deployment of AFCI facilities, cost tradeoffs to meet nuclear non-proliferation requirements, and international nuclear facility supply/demand analysis. The economic analysis will be performed using two cost models. VISION.ECON will be used to evaluate and compare costs under dynamic conditions, consistent with the cases and analysis performed by the AFCI Systems Analysis team. Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) will provide static (snapshot-in-time) cost analysis and will provide a check on the dynamic results. In future analysis, additional AFCI measures may be developed to show the value of AFCI in closing the fuel cycle. Comparisons can show AFCI in terms of reduced global proliferation (e.g., reduction in enrichment), greater sustainability through preservation of a natural resource (e.g., reduction in uranium ore depletion), value from

  7. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

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

    International Nuclear Information System (INIS)

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

  9. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  10. 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.

  11. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    Energy Technology Data Exchange (ETDEWEB)

    Heinz Pitsch

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high-fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation, a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet tranformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  12. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    Energy Technology Data Exchange (ETDEWEB)

    Pitsch, Heinz

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation; a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet transformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  13. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  14. A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

    International Nuclear Information System (INIS)

    The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT planned

  15. Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

    International Nuclear Information System (INIS)

    The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs

  16. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    CERN Document Server

    Sinev, V V

    2009-01-01

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  17. A review of the prospects for fusion breeding of fissile material

    International Nuclear Information System (INIS)

    This report is the result of an eight month study by the AECL Fusion Status Study Group. The objectives of this study were to review the current status of fusion research, to evaluate the neutronic performance of various fusion-breeder systems, and to assess the economic and technological outlook for the fusion breeder as a source of fissile material to support CANDU reactors operating on the thorium fuel cycle

  18. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  19. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  20. Masters Study in Advanced Energy and Fuels Management

    Energy Technology Data Exchange (ETDEWEB)

    Mondal, Kanchan [Southern Illinois Univ., Carbondale, IL (United States)

    2014-12-08

    There are currently three key drivers for the US energy sector a) increasing energy demand and b) environmental stewardship in energy production for sustainability and c) general public and governmental desire for domestic resources. These drivers are also true for energy nation globally. As a result, this sector is rapidly diversifying to alternate sources that would supplement or replace fossil fuels. These changes have created a need for a highly trained workforce with a the understanding of both conventional and emerging energy resources and technology to lead and facilitate the reinvention of the US energy production, rational deployment of alternate energy technologies based on scientific and business criteria while invigorating the overall economy. In addition, the current trends focus on the the need of Science, Technology, Engineering and Math (STEM) graduate education to move beyond academia and be more responsive to the workforce needs of businesses and the industry. The SIUC PSM in Advanced Energy and Fuels Management (AEFM) program was developed in response to the industries stated need for employees who combine technical competencies and workforce skills similar to all PSM degree programs. The SIUC AEFM program was designed to provide the STEM graduates with advanced technical training in energy resources and technology while simultaneously equipping them with the business management skills required by professional employers in the energy sector. Technical training include core skills in energy resources, technology and management for both conventional and emerging energy technologies. Business skills training include financial, personnel and project management. A capstone internship is also built into the program to train students such that they are acclimatized to the real world scenarios in research laboratories, in energy companies and in government agencies. The current curriculum in the SIUC AEFM will help fill the need for training both recent

  1. Foresight Study on Advanced Conversion Technologies of Fossil Fuels

    International Nuclear Information System (INIS)

    The Observatorio de Prospectiva Tecnologica Industrial (OPTI) is a Foundation supported by the Ministry of Industry and Energy, (MINER) and has as main objective to provide a basic information and knowledge on technology evolution. This information will be accessible to the Administration and to the Companies and can be taking into account in planning and decision making of technology policies. Ciemat is member of OPTI and is the organism in charge of the actions in the Energy sector. CIEMAT has the responsibility on the realisation of the sector studies to get in three years (1998 to 2001) a foresight vision of the critical technology topics. The OPTI integrated strategic plan undertake the analysis of other seven technology sectors, with the same criteria on methodological aspects. Delphi method was used for the realization of the studies. It consisted of a survey conducted in two rounds using a questionnaire to check the experts opinion. The time frame of the studies was defined from 1999 to 2015. The study presented in this document has been performed by CIEMAT in the second stage of the OPTI activities. The main goal behind this study is to identify the advanced clean and efficient technologies for the conversion of fossil fuels to promote in our country. The questionnaire was addressed to 250 experts and the response rate was about the 37%, ratifying the final results. The spanish position and the barriers for the development of each technology has been determined and also the recommended measures to facilitate their performance in the future. This basic information is consider of main interest, taking in account the actual energetic situation with a foreseeable demand increase and fossil fuels dependence. (Author) 17 refs

  2. A review on the development of the advanced fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author).

  3. Administrative Co-ordination of Fissile Material Management and Accounting in the U.K.A.E.A

    International Nuclear Information System (INIS)

    The Authority are engaged as suppliers in fissile material production, distribution, recycle and reprocessing. As consumers, the Authority require fissile material for power reactors, a variety of prototypes, MTRs, zero-energy facilities and fuel development projects; and for other experimental and research purposes in laboratory quantities. Executive responsibility for these activities lies with the four Groups through which the Authority discharge these functions. It has been found useful to keep these activities under review in specialized inter-Group Committees, with a common secretariat. These Committees: (a) study all projects all proposals or work involving significant quantities of fissile material (plutonium and enriched uranium, other than natural U or U depleted in 235U) in the light of expected supplies over a number of years from all sources, including new production, scrap recovery and imports; and all uses including burn-up, losses and exports; (b) recommend the optimum allocation of specific amounts for approved purposes in relation to other calls upon available supplies, and having regard to the economic issues involved; (c) record and progress all approved allocations, and examine the nature, amount and purpose of all existing stockholdings in relation to current policies and objectives; (d) record and study all losses of fissile material during fabrication or other processing and the measures taken to reduce them; (e) assist in developing procedures and incentives to ensure that material is used economically and returned promptly. Each Group has considerable autonomy in its day-to-day use of fissile material. The administrative machinery described above provides a means by which the Authority’s scientists, engineers, accountants and administrators concerned with fissile material problems can operate collectively in a common frame of reference with a minimum of paperwork. The paper is illustrated with a simplified flowsheet of the main flows of

  4. Recent Advances in Enzymatic Fuel Cells: Experiments and Modeling

    Directory of Open Access Journals (Sweden)

    Ivan Ivanov

    2010-04-01

    Full Text Available Enzymatic fuel cells convert the chemical energy of biofuels into electrical energy. Unlike traditional fuel cell types, which are mainly based on metal catalysts, the enzymatic fuel cells employ enzymes as catalysts. This fuel cell type can be used as an implantable power source for a variety of medical devices used in modern medicine to administer drugs, treat ailments and monitor bodily functions. Some advantages in comparison to conventional fuel cells include a simple fuel cell design and lower cost of the main fuel cell components, however they suffer from severe kinetic limitations mainly due to inefficiency in electron transfer between the enzyme and the electrode surface. In this review article, the major research activities concerned with the enzymatic fuel cells (anode and cathode development, system design, modeling by highlighting the current problems (low cell voltage, low current density, stability will be presented.

  5. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    International Nuclear Information System (INIS)

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  6. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  7. Fuel economy screening study of advanced automotive gas turbine engines

    Science.gov (United States)

    Klann, J. L.

    1980-01-01

    Fuel economy potentials were calculated and compared among ten turbomachinery configurations. All gas turbine engines were evaluated with a continuously variable transmission in a 1978 compact car. A reference fuel economy was calculated for the car with its conventional spark ignition piston engine and three speed automatic transmission. Two promising engine/transmission combinations, using gasoline, had 55 to 60 percent gains over the reference fuel economy. Fuel economy sensitivities to engine design parameter changes were also calculated for these two combinations.

  8. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Spanish version); Clean Cities, Energy Efficiency & Renewable Energy (EERE)

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Erik

    2015-06-01

    Powering commercial lawn equipment with alternative fuels or advanced engine technology is an effective way to reduce U.S. dependence on petroleum, reduce harmful emissions, and lessen the environmental impacts of commercial lawn mowing. Numerous alternative fuel and fuel-efficient advanced technology mowers are available. Owners turn to these mowers because they may save on fuel and maintenance costs, extend mower life, reduce fuel spillage and fuel theft, and demonstrate their commitment to sustainability.

  9. 49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

    Science.gov (United States)

    2010-10-01

    ... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the...

  10. Update on Monitoring Technologies for International Safeguards and Fissile Material Verification

    International Nuclear Information System (INIS)

    Monitoring technologies are playing an increasingly important part in international safeguards and fissile material verification. The developments reduce the time an inspector must spend at a site while assuring continuity of knowledge. Monitoring technologies' continued development has produced new seal systems and integrated video surveillance advances under consideration for Trilateral Initiative use. This paper will present recent developments for monitoring systems at Embalse, Argentina, VNHEF, Sarov, Russian, and Savannah River Site, Aiken, South Carolina

  11. Criteria and Evaluation for the Storage of Fissile Material in a Large and Varied Reactor Research and Development Programme

    International Nuclear Information System (INIS)

    Precise analysis of the neutron interaction between fissile material containers is possible through the application of Monte Carlo techniques such as the GEM code of the U.K.A.E.A. or O5R developed by Oak Ridge National Laboratory. While this is appropriate for well defined and inflexible arrays, many problems of practical materials storage do not require this rigor, nor are many materials storage configurations sufficiently well defined to permit full advantage to be derived from such treatment. An analysis which is amenable to slide rule calculation has been found sufficient for most of the problems that arise in a laboratory which has a large inventory of fissile material in the many forms required for a large, extensive, and varied reactor research and development programme (including fuels and materials development). This presentation is directed toward the nuclear safety specialist who must, with limited support facilities, derive criteria for the safe storage of fissile material without undue economic penalty. (author)

  12. Screening of advanced cladding materials and UN–U3Si5 fuel

    International Nuclear Information System (INIS)

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U3Si5 fuels have the potential to exhibit reactor physics and fuel management performance similar to UO2. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN–U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN–U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels

  13. Fission dynamics with systems of intermediate fissility

    Indian Academy of Sciences (India)

    E Vardaci; A Di Nitto; P N Nadtochy; A Brondi; G La Rana; R Moro; M Cinausero; G Prete; N Gelli; E M Kozulin; G N Knyazheva; I M Itkis

    2015-08-01

    A 4 light charged particle spectrometer, called 8 LP, is in operation at the Laboratori Nazionali di Legnaro, Italy, for studying reaction mechanisms in low-energy heavy-ion reactions. Besides about 300 telescopes to detect light charged particles, the spectrometer is also equipped with an anular PPAC system to detect evaporation residues and a two-arm time-of-flight spectrometer to detect fission fragments. The spectrometer has been used in several fission dynamics studies using as a probe light charged particles in the fission and evaporation residues (ER) channels. This paper proposes a journey within some open questions about the fission dynamics and a review of the main results concerning nuclear dissipation and fission time-scale obtained from several of these studies. In particular, the advantages of using systems of intermediate fissility will be discussed.

  14. Measurements of inventories with mixed fissile materials

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Krick, M.S.; Kelley, T.; Schneider, C.M. [and others

    1997-11-01

    An inventory with a large number of diverse items containing mixtures of uranium and plutonium has been measured with two nondestructive assay (NDA) instruments used in four modes. A segmented gamma scanner (SGS) was used to find the number of cans and the positions of the fissile materials by scanning each item in front of a transmissions source; at each position, uranium and plutonium isotopics were measured with the passive gamma rays emitted. A shuffler was then used in both the passive and active modes to measure the masses of the two elements. The measured masses for the inventory items were generally in agreement with the declared values, but anomalies were identified for a small fraction of the inventory.

  15. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L.W.

    1996-01-08

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ``excess`` nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist.

  16. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    International Nuclear Information System (INIS)

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ''excess'' nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist

  17. Fissile and non-fissile element separation in concrete radioactive waste drums using the SIMPHONIE method

    International Nuclear Information System (INIS)

    The simultaneous photon and neutron interrogation experiment (SIMPHONIE) method, applied to radioactive waste drum characterization, has already been treated in [F. Jallu, A. Lyoussi, C. Passard, E. Payan, H. Recroix, G. Nurdin, A. Buisson, J. Allano, Nucl. Instr. and Meth. B 170 (2000) 489]. First experimental results carried out with U and Pu bare samples were presented, that showed the feasibility of quantifying fissile (235U, 239,241Pu, ...) and non-fissile (234,236,238U, 238,240Pu, ...) elements separately in only one measurement, using both active neutron interrogation and induced photofission interrogation techniques simultaneously. This paper presents new experimental results carried out with U samples embedded in a concrete matrix. These results have been obtained using the DGA/ETCA MiniLinatron pulsed linear electron accelerator located at Arcueil, France. Mass detection limits of less than 2 g of matter have been obtained with the preliminary setup used in these experiments

  18. An implicit solution framework for reactor fuel performance simulation

    International Nuclear Information System (INIS)

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding that surrounds the pellets. An important design goal for a fuel is to maximize the life of the cladding thereby allowing the fuel to remain in the reactor for a longer period of time to achieve higher degrees of burnup. This presentation examines various mathematical and computational issues that impact the modeling of the thermomechanical response of reactor fuel, and are thus important to the development of INL's fuel performance analysis code, BISON. The code employs advanced methods for solving coupled partial differential equation systems that describe multidimensional fuel thermomechanics, heat generation, and transport within the fuel

  19. Fuel fragmentation model advances using TEXAS-V

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

    1998-01-01

    Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

  20. Advanced nuclear fuel study for the utilization of carbon-coated

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee Unviersity, Seoul (Korea)

    1998-03-01

    Advanced nuclear fuel design of carbon coated fuel particles(UCO fuel) was suggested to the current PWRs. Nuclear feasibility studying was forformed for the double heterogeneous UCO fuel by CASMO-3. UCO fuel showed nuclear feasibility when they were packed in the Ulchin3/4 fuel assembly. Nuclear safety was evaluated for the UCO fuel by FTC an dMTC, which had enough safety at operating condition. The average fuel temperature compared with conventional oxide fuel at hot full power condition was reduced by 150 deg K, which was caused by high conductivity of carbon matrix. A core design, used UCO fuel, was possible for same forformance with Ulchin3/4. But, UCO fuel enrichment exceed the PWR fuel enrichment limit 5w/o. Cycle length of UCO duel core was shortened by 90 EFPD satisfied with enrichment limit and thermal power. It is not good for using UCO fuel in PWRs in respect of fuel costs. (author). 19 refs., 71 figs., 25 tabs.

  1. Evolution of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear fuel is the primary energy source for sustaining the nuclear fission chain reactions in a reactor. The fuels in the reactor cores are exposed to highly aggressive environment and varieties of advanced fuel materials with improved nuclear properties are continuously being developed to have optimum performance in the existing core conditions. Fabrications of varieties of nuclear fuels used in diverse forms of reactors are mainly based on two naturally occurring nuclear source elements, uranium as fissile 235U and fertile 238U, and thorium as fertile 232Th species. The two metals in the forms of alloys with specific elements, ceramic oxides like MOX and ceramic non-oxide as mixed carbide and nitride with suitable nuclear properties like higher metal density, thermal conductivity, etc. are used as fuels in different reactor designs. In addition, efficiency of various advanced fuels in the forms of dispersion, molten salt and other types are also under investigations. The countries which have large deposits of thorium but limited reserves of uranium, are trying to give special impetus on the development of thorium-based fuels for both thermal and fast reactors in harnessing nuclear energy for peaceful uses of atomic energy. (author)

  2. Development of Advanced Hydrocarbon Fuels at Marshall Space Flight Center

    Science.gov (United States)

    Bai, S. D.; Dumbacher, P.; Cole, J. W.

    2002-01-01

    This was a small-scale, hot-fire test series to make initial measurements of performance differences of five new liquid fuels relative to rocket propellant-1 (RP-1). The program was part of a high-energy-density materials development at Marshall Space Flight Center (MSFC), and the fuels tested were quadricyclane, 1-7 octodiyne, AFRL-1, biclopropylidene, and competitive impulse noncarcinogenic hypergol (CINCH) (di-methyl-aminoethyl-azide). All tests were conducted at MSFC. The first four fuels were provided by the U.S. Air Force Research Laboratory (AFRL), Edwards Air Force Base, CA. The U.S. Army, Redstone Arsenal, Huntsville, AL, provided the CINCH. The data recorded in all hot-fire tests were used to calculate specific impulse and characteristic exhaust velocity for each fuel, then compared to RP-1 at the same conditions. This was not an exhaustive study, comparing each fuel to RP-1 at an array of mixture ratios, nor did it include important fuel parameters, such as fuel handling or long-term storage. The test hardware was designed for liquid oxygen (lox)/RP-1, then modified for gaseous oxygen/RP-1 to avoid two-phase lox at very small flow rates. All fuels were tested using the same thruster/injector combination designed for RP-1. The results of this test will be used to determine which fuels will be tested in future test programs.

  3. Advanced nuclear fuel cycles - Main challenges and strategic choices

    International Nuclear Information System (INIS)

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

  4. Advanced Composite Bipolar Plate for Unitized Regenerative Fuel Cell/Electrolyzer Systems Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Development of an advanced composite bipolar plate is proposed for a unitized regenerative fuel cell and electrolyzer system that operates on pure feed streams...

  5. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  6. The U.S. Advanced Fuel Cycle Initiative: Development of separations technologies

    International Nuclear Information System (INIS)

    Spent nuclear fuel from 103 operating U.S. commercial nuclear power reactors is accumulating at a rate of about 2,000 metric tons per year. At this rate, the legislated capacity of the Yucca Mountain geologic repository (63,000 tons of commercial spent fuel) will be exceeded by 2015. Accordingly, the U.S. Department of Energy has instituted a new program, the Advanced Fuel Cycle Initiative, which is intended to provide the technologies necessary for the economical and environmentally sound processing of spent fuel. The goal of this technology development program is to preclude or significantly delay the need for a second geologic repository. Separations technologies are being developed that will support the processing of commercial spent fuel as well as the spent fuel arising from the operation of future advanced reactors

  7. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  8. The role of advanced calculation and simulation tools in the evolution of fuel

    International Nuclear Information System (INIS)

    This article is focused on the role of the advanced calculation/simulation tools on the development of the fuel designs as well as in the assessment of the effect of the changes in the operation. With this purpose, the article describes and shows some examples of the use by ENUSA of some of these tools in the fuel engineering. To conclude, the future on the evolution of the advanced tools is also presented. (Author)

  9. Advanced fuel developments for an industrial accelerator driven system prototype

    Energy Technology Data Exchange (ETDEWEB)

    Delage, Fabienne; Ottaviani, Jean Pierre [Commissariat a l' Energie Atomique CEA (France); Fernandez-Carretero, Asuncion; Staicu, Dragos [JRC-ITU (Germany); Boccaccini, Claudia-Matzerath; Chen, Xue-Nong; Mascheck, Werner; Rineiski, Andrei [Forschungszentrum Karlsruhe - FZK (Germany); D' Agata, Elio [JRC-IE (Netherlands); Klaassen, Frodo [NRG, PO Box 25, NL-1755 ZG Petten (Netherlands); Sobolev, Vitaly [SCK-CEN (Belgium); Wallenius, Janne [KTH Royal Institute of Technology (Sweden); Abram, T. [National Nuclear Laboratory - NNL (United Kingdom)

    2009-06-15

    Fuel to be used in an Accelerator Driven System (ADS) for transmutation in a fast spectrum, can be described as a highly innovative concept in comparison with fuels used in critical cores. ADS fuel is not fertile, so as to improve the transmutation performance. It necessarily contains a high concentration ({approx}50%) of minor actinides and plutonium. This unusual fuel composition results in high gamma and neutron emissions during its fabrication, as well as degraded core performance. So, an optimal ADS fuel is based on finding the best compromise between thermal, mechanical, chemical, neutronic and technological constraints. CERCER and CERMET composite fuels consisting of particles of (Pu,MA)O{sub 2} phases dispersed in a magnesia or molybdenum matrix are under investigation within the frame of the ongoing European Integrated Project EUROTRANS (European Research programme for Transmutation) which aims at performing a conceptual design of a 400 MWth transmuter: the European Facility for Industrial Transmutation (EFIT). Performances and safety of EFIT cores loaded with CERCER and CERMET fuels have been evaluated. Out-of-pile and in-pile experiments are carried out to gain knowledge on the properties and the behaviour of these fuels. The current paper gives an overview of the work progress. (authors)

  10. Development of advanced combustion technology for biomass fuels

    Energy Technology Data Exchange (ETDEWEB)

    Douglas, M.A.; Wong, S.; Jones, A.K

    1994-04-01

    An innovative inclined dual-grate concept for drying wood refuse fuels was evaluated using a pilot plant in 1:1 scale clod flow studies. The dual-grate concept adapts the mass burn technique widely used for incinerating municipal solid waste and applies it to wet woodwaste/sludge mixtures. Screw feeders admit the mixtures onto a steeply inclined, bare-tube, watercooled grate that may have two or more panels at different slopes to control the depth of the fuel bed. Hot undergrate air dries the fuel bed and contributes to a steady movement of the fuel down each inclined panel. Devolatilization, ignition, and some burning of the fuel occurs at the lowest portion of the drying grate, which discharges onto a burnout grate where combustion is completed and from which ash is continuously discharged. Nine wood waste feedstocks were tested in the pilot unit to determine the effects of sludge, moisture, and sawdust or shavings additions to a base fuel on fuel bed velocity for various grate angles and undergrate air flows. Optimal angles for the upper and lower grates appear to be ca 35{degree} and 25{degree} respectively with fuel beds between 4 and 10 in. and plenum air pressures between 3 and 9 in. water gauge. The cold tests indicated that the dual grate design has excellent potential as a drying grate for wood refuse containing large amounts of moisture and deinking sludge. Pilot plant operation can be improved with minor design modifications. 42 figs., 6 tabs.

  11. An examination of the elastic structural response of the Advanced Neutron Source fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Swinson, W.F.; Luttrell, C.R.; Yahr, G.T.

    1994-09-01

    Procedures for evaluating the elastic structural response of the Advanced Neutron Source (ANS) fuel plates to coolant flow and to temperature variations are presented in this report. Calculations are made that predict the maximum deflection and the maximum stress for a representative plate from the upper and from the lower fuel elements.

  12. Fissile material measurements using the differential die-away self interrogation technique

    Energy Technology Data Exchange (ETDEWEB)

    Schear, Melissa A [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Evans, Louise G [Los Alamos National Laboratory; Lee, S Y [Los Alamos National Laboratory

    2010-01-01

    Currently, there is substantial research effort focused on quantifying plutonium (Pu) mass in spent fuel using non-destructive assay (NDA) techniques. Of the several techniques being investigated for this purpose, Differential Die-Away Self-Interrogation (DDSI) is a recently proposed, neutron-based NDA technique capable of quantifying the total fissile content in an assembly. Unlike the conventional Differential Die-Away (DDA) technique, DOSI does not require an external neutron source for sample interrogation, but rather, uses the spontaneous fission neutrons originating from {sup 244}Cm within the spent fuel for self-interrogation. The essence of the technique lies in the time separation between the detection of spontaneous fission neutrons from {sup 244}Cm and the detection of induced fission neutrons at a later time. The DDSI detector design imposes this time separation by optimizing the die-away times ({tau}) of the detector and sample interrogation regions to obtain an early and late neutron distribution respectively. The ratio of the count rates in the late gate to the early gate for singles, doubles, and triples is directly proportional to the fissile content present in the sample, which has already been demonstrated for simplified fuel cases using the Monte Carlo N-Particle eXtended (MCNPX) code. The current work applies the DDSI concept to more complex samples, specifically spent Pressurized Water Reactor (PWR) assemblies with varying isotopics resulting from a range of initial enrichment, bumup, and cooling time. We assess the feasibility of using the late gate to early gate ratio as a reliable indicator of overall fissile mass for a range of assemblies by defining a {sup 239}Pu effective mass which indicates the mass of {sup 239}Pu that would yield the same DDSI signal as the combined mass of major fissile isotopes present in the sample. This work is important for assessing the individual capability of the DDSI instrument in quantifying fissile mass in

  13. Advanced anodes for high-temperature fuel cells

    DEFF Research Database (Denmark)

    Atkinson, A.; Barnett, S.; Gorte, R.J.;

    2004-01-01

    Fuel cells will undoubtedly find widespread use in this new millennium in the conversion of chemical to electrical energy, as they offer very high efficiencies and have unique scalability in electricity-generation applications. The solid-oxide fuel cell (SOFC) is one of the most exciting...... of these energy technologies; it is an all-ceramic device that operates at temperatures in the range 500-1,000degreesC. The SOFC offers certain advantages over lower temperature fuel cells, notably its ability to use carbon monoxide as a fuel rather than being poisoned by it, and the availability of high......-grade exhaust heat for combined heat and power, or combined cycle gas-turbine applications. Although cost is clearly the most important barrier to widespread SOFC implementation, perhaps the most important technical barriers currently being addressed relate to the electrodes, particularly the fuel electrode...

  14. Advances in Materials and System Technology for Portable Fuel Cells

    Science.gov (United States)

    Narayanan, Sekharipuram R.

    2007-01-01

    This viewgraph presentation describes the materials and systems engineering used for portable fuel cells. The contents include: 1) Portable Power; 2) Technology Solution; 3) Portable Hydrogen Systems; 4) Direct Methanol Fuel Cell; 5) Direct Methanol Fuel Cell System Concept; 6) Overview of DMFC R&D at JPL; 7) 300-Watt Portable Fuel Cell for Army Applications; 8) DMFC units from Smart Fuel Cell Inc, Germany; 9) DMFC Status and Prospects; 10) Challenges; 11) Rapid Screening of Well-Controlled Catalyst Compositions; 12) Screening of Ni-Zr-Pt-Ru alloys; 13) Issues with New Membranes; 14) Membranes With Reduced Methanol Crossover; 15) Stacks; 16) Hybrid DMFC System; 17) Small Compact Systems; 18) Durability; and 19) Stack and System Parameters for Various Applications.

  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. The processes for fissile material flux management: theory vs practice

    International Nuclear Information System (INIS)

    Bel-V has carried out a thorough control of some licensees in order to check if the official accountability of fissile material matched the experimental fissile material fluxes in the processes. The following approach was adopted: 1) Understand how the licensee has kept count of the official accountability over the years (it must be noted that the past method used was evaluated and recognized by the regulator); 2) Observe and get an in-depth understanding of the processes carried out by the licensee; 3) Compare these processes with the information (yields, chemical reactions, byproducts formation...) published in scientific literature; 4) Localize parts of the installation or equipment where an accumulation of fissile material could take place; 5) Ask the licensee for a sample analysis in parts of the equipment where the localization study showed a risk of accumulation or loss of fissile material; 6) Compare the analysis results with the official accountability values; 7) If applicable, suggest any process modification in order to improve the recovery of fissile material; 8) In the future, eventually estimate the potential necessary corrections of the official accountability of fissile material (administrative task). Bel-V has not detected any major problems that necessitate immediate action. However, this study has identified parts of equipment in some facilities where the accumulation rate of fissile material is higher than expected. On the other hand, experimental analyzes have excluded some places where fissile material was supposed to be prone to accumulation. Finally the licensees and Bel-V have localized places where chemical side reactions or very slow kinetic reactions have altered the molecular form of the fissile material. The article is followed by the slides of the presentation

  17. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-01

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  18. ACSEPT, Toward the Future Demonstration of Advanced Fuel Treatments

    Energy Technology Data Exchange (ETDEWEB)

    Bourg, Stephane; Hill, Clement [CEA/DEN/MAR/DRCP, Marcoule, BP17171, 30207 Bagnols/ceze (France); Caravaca, Concha [CIEMAT (Spain); Ekberg, Christian [CHALMERS University (Sweden); Rhodes, Chris [Nuclear National Laboratory (United Kingdom)

    2009-06-15

    Actinide recycling by separation and transmutation is considered worldwide and particularly in several European countries as one of the most promising strategies to reduce the inventory of radioactive waste and to optimize the use of natural resources, thus contributing to making nuclear energy sustainable. In accordance with the Strategic Research Agenda (SRA) of the Sustainable Nuclear Energy Technology Platform (SNE-TP), the timelines of the FP7-EURATOM project ACSEPT (2008-2012) should allow the offering of technical solutions in terms of advanced closed fuel cycle technologies including the recycling of actinides and that may be reviewed by Governments, European utilities as well as Technology Providers at the time horizon 2012. By joining in its consortium 34 partners from 12 European countries plus Australia and Japan, ACSEPT is thus an essential contribution to the demonstration, in the long term, of the potential benefits of actinide recycling. To succeed, ACSEPT is organized into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimize and select the most promising ones dedicated either to actinide partitioning or to grouped actinide separation. A substantial review was undertaken either to be sure that the right molecule families are being studied, or, on the contrary, to identify new candidates. After 18 months, results of the first hot tests should allow the validation of some process options. In addition, the first results on dissolution studies will be available as well as the progress in conversion techniques. (ii) Concerning pyrochemical separation processes, ACSEPT is focused on the enhancement of the two reference cores of process selected within EUROPART with specific attention to the exhaustive electrolysis in molten chloride (quantitative recovery of the actinides with the lowest amount of fission products) and to actinide back-extraction from an An-Al alloy. R and D efforts are also

  19. γ-ray self-absorption of cylindrical fissile material

    Institute of Scientific and Technical Information of China (English)

    HUANG Yong-Yi; CHENG Yi-Ying; TIAN Dong-Feng; LU Fu-Quan; YANG Fu-Jia

    2005-01-01

    The self-absorption of γ-ray emitted from cylindrical fissile materials, such as 235U and 239Pu, does not possess spherical symmetry. The analytical formulae of self-absorption for γ-ray throughout the cylinder have been obtained. The intensity of γ-ray is a function of γ-ray outgoing directions and cylindrical configurations, accordingly one can acquire the information about geometrical configuration of cylindrical fissile materials through multi-location measurements. Further more, the method is given in this article. The result can be applied to the fissile material safeguard, such as nuclear monitoring and verifying.

  20. A Review of Thorium Utilization as an option for Advanced Fuel Cycle-Potential Option for Brazil in the Future

    International Nuclear Information System (INIS)

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO2, as will be reviewed in this paper, and therefore RandD programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle

  1. Organic coal-water fuel: Problems and advances (Review)

    Science.gov (United States)

    Glushkov, D. O.; Strizhak, P. A.; Chernetskii, M. Yu.

    2016-10-01

    The study results of ignition of organic coal-water fuel (OCWF) compositions were considered. The main problems associated with investigation of these processes were identified. Historical perspectives of the development of coal-water composite fuel technologies in Russia and worldwide are presented. The advantages of the OCWF use as a power-plant fuel in comparison with the common coal-water fuels (CWF) were emphasized. The factors (component ratio, grinding degree of solid (coal) component, limiting temperature of oxidizer, properties of liquid and solid components, procedure and time of suspension preparation, etc.) affecting inertia and stability of the ignition processes of suspensions based on the products of coaland oil processing (coals of various types and metamorphism degree, filter cakes, waste motor, transformer, and turbine oils, water-oil emulsions, fuel-oil, etc.) were analyzed. The promising directions for the development of modern notions on the OCWF ignition processes were determined. The main reasons limiting active application of the OCWF in power generation were identified. Characteristics of ignition and combustion of coal-water and organic coal-water slurry fuels were compared. The effect of water in the composite coal fuels on the energy characteristics of their ignition and combustion, as well as ecological features of these processes, were elucidated. The current problems associated with pulverization of composite coal fuels in power plants, as well as the effect of characteristics of the pulverization process on the combustion parameters of fuel, were considered. The problems hindering the development of models of ignition and combustion of OCWF were analyzed. It was established that the main one was the lack of reliable experimental data on the processes of heating, evaporation, ignition, and combustion of OCWF droplets. It was concluded that the use of high-speed video recording systems and low-inertia sensors of temperature and gas

  2. Operational experiences in MOX fuel fabrication for the FUGEN advanced thermal reactor

    International Nuclear Information System (INIS)

    The Japan Nuclear Cycle Development Institute, JNC, has fabrication the MOX fuel for the Advanced Thermal Reactor, ATR, ''FUGEN'' in the Plutonium Fuel Fabrication Facility, PFFF, since 1974. For these 25 years, the MOX fuel fabrication has progressed in stable manner after overcoming several problems at the start up of FUGEN fuel fabrication. Through the experience, improvements on process equipment and conditions have been taken place to achieve efficient MOX fuel fabrication on an engineering scale as 10 tons MOX per year. Main features of current fabrication process are digested as one step blending with ball milling, pelletizing without granulation and sintering with batch type furnaces. This fabrication process has been demonstrated and confirmed to be applicable techniques for the MOX fuel fabrication on this scale. This paper discusses the FUGEN fuel fabrication focused on the MOX pellet fabrication with operational experiences and improvements to the process. (author)

  3. Advanced Space Power Systems (ASPS): Regenerative Fuel Cells (RFC) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the regenerative fuel cell project element is to develop power and energy storage technologies that enable new capabilities for future human space...

  4. 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.

  5. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  6. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  7. Advanced multiphysics coupling for LWR fuel performance analysis

    International Nuclear Information System (INIS)

    Highlights: • Overviews the BISON nuclear fuel performance analysis code. • Discusses loose and tight coupling approaches applied to thermomechanics. • Outlines coupling to the DeCART neutronics code. • Demonstrates multiscale coupling using MARMOT. - Abstract: Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution

  8. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  9. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: First Results Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2011-03-01

    This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This report provides the early data results and implementation experience of the AT fuel cell bus since it was placed in service.

  10. Advanced diagnostics in oxy-fuel combustion processes

    Energy Technology Data Exchange (ETDEWEB)

    Brix, J.; Clausen, Soennik; Degn Jensen, A. (Technical Univ. of Denmark. CHEC Research Centre, Kgs. Lyngby (Denmark)); Boeg Toftegaard, M. (DONG Energy Power, Hvidovre (Denmark))

    2012-07-01

    This report sums up the findings in PSO-project 010069, ''Advanced Diagnostics in Oxy-Fuel Combustion Processes''. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory scale fixed bed reactor. The results obtained in the swirl burner have proved the FTIR method as a valuable technique for gas phase temperature measurements. When its efficacy is evaluated against traditional thermocouple measurements, two cases, with and without probe beam stop, must however be treated separately. When the FTIR probe is operated with the purpose of gas phase concentration measurements the probe needs to operate with a beam stop mounted in front of it. With this beam stop in place it was shown that the measured gas phase temperature was affected by cooling, induced by the cooled beam stop. Hence, for a more accurate determination of gas phase temperatures the probe needed to operate without the beam stop. When this was the case, the FTIR probe showed superior to traditional temperature measurements using a thermocouple as it could measure the fast temperature fluctuations. With the beam stop in place the efficacy of the FTIR probe for gas temperature determination was comparable to the use of a traditional thermocouple. The evaluation of the FTIR technique regarding estimation of gas phase concentrations of H{sub 2}O, CO{sub 2} and CO showed that the method is reliable though it cannot be stated as particularly accurate. The accuracy of the method is dependent on the similarity of the reference emission spectra of the gases with those obtained in the experiments, as the transmittance intensity is not a linear function of concentration. The length of the optical path also affects the steadiness of the measurements. The length of the optical path is difficult to adjust on the small scales that are the focus of this work. However

  11. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  12. Gasoline Ultra Efficient Fuel Vehicle with Advanced Low Temperature Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Confer, Keith

    2014-09-30

    The objective of this program was to develop, implement and demonstrate fuel consumption reduction technologies which are focused on reduction of friction and parasitic losses and on the improvement of thermal efficiency from in-cylinder combustion. The program was executed in two phases. The conclusion of each phase was marked by an on-vehicle technology demonstration. Phase I concentrated on short term goals to achieve technologies to reduce friction and parasitic losses. The duration of Phase I was approximately two years and the target fuel economy improvement over the baseline was 20% for the Phase I demonstration. Phase II was focused on the development and demonstration of a breakthrough low temperature combustion process called Gasoline Direct- Injection Compression Ignition (GDCI). The duration of Phase II was approximately four years and the targeted fuel economy improvement was 35% over the baseline for the Phase II demonstration vehicle. The targeted tailpipe emissions for this demonstration were Tier 2 Bin 2 emissions standards.

  13. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    International Nuclear Information System (INIS)

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  14. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    D.E. Clark; D.C. Folz

    2010-08-29

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  15. Advances in PEM fuel cells with CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Robalinho, Eric; Cunha, Edgar Ferrari da; Zararya, Ahmed; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], Email: eric@ipen.br; Cekinski, Efrain [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil)

    2010-07-01

    This paper presents some applications of computational fluid dynamics techniques in the optimization of Proton Exchange Membrane Fuel Cell (PEMFC) designs. The results concern: modeling of gas distribution channels, the study for both porous anode and cathode and the three-dimensional modeling of a partial geometry layer containing catalytic Gas Diffusion Layers (GDL) and membrane. Numerical results of the simulations of graphite plates flow channels, using ethanol as fuel, are also presented. Some experimental results are compared to the corresponding numerical ones for several cases, demonstrating the importance and usefulness of this computational tool. (author)

  16. Constituent Redistribution in U-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Matthews, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-30

    Previous work done by Galloway, et. al. [1] on EBR-II ternary (U-Pu-Zr) fuel constituent redistribution yielded accurate simulation data for the limited data sets of Zr redistribution. The data sets included EPMA scans of two different irradiated rods. First, T179 which was irradiated to 1.9 at% burnup was analyzed. Second, DP16 which was irradiated to 11 at% burnup was analyzed. One set of parameters that most accurately represented the zirconium profiles for both experiments was determined. Since the binary fuel (U-Zr) has previously been used as the driver fuel for sodium fast reactors (SFR) as well as being the likely driver fuel if a new SFR is constructed, this same process has been initiated on the binary fuel form. From limited binary EPMA scans as well as other fuel characterization techniques, it has been observed that zirconium redistribution also occurs in the binary fuel, albeit at a reduced rate compared to observation in the ternary fuel, as noted by Kim et. al. in [2]. While the rate of redistribution has been observed to be slower, numerous metallographs of U-Zr fuel, such as the one shown in Figure 1, show distinct zone formations.

  17. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  18. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  20. Advanced energy analysis of high temperature fuel cell systems

    NARCIS (Netherlands)

    De Groot, A.

    2004-01-01

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

  1. 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.

  2. PHWR advanced fuel R and D for the 21st Century in Korea

    International Nuclear Information System (INIS)

    As the first CANDU-PHWR nuclear power plant in Korea, Wolsong Unit 1 has been successfully operated since 1983. The CANDU installed electric-generation capacity was about 50 % of the installed electric-generation capacity of nuclear power plants in Korea in 1983 but then decreased to less than 10 % of the total installed nuclear electric-generation capacity by 1996. This CANDU installed electric-generation capacity has recovered to about 20 % of the total installed nuclear electric-generation capacity in 1999, because Wolsong Units 2, 3 and 4 have been placed into commercial operation in 1997, 1998 and 1999, respectively. This indicates that CANDU reactors are not the majority of nuclear power plants in Korea. Since the period of the late 1970s, nuclear fuel design and fabrication technologies have been engaged as one of the important R and D activities in Korea. As one of the early R and D activities leading to nuclear power industrialization in Korea, the project to develop the design and fabrication technology of CANDU-6 37-element fuel had been successfully carried out from 1981 to 1987 by KAERI. Just after the successful completion of the 37-element fuel R and D, KAERI has developed a CANDU-6 advanced fuel. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology. The CANFLEX and DUPIC R and D programs have been conducted under Korea's Nuclear Energy R and D Project as national mid- and long-term programs since 1992. As the second of the CANDU R and D products in Korea, the CANFLEX-NU fuel has been jointly developed by KAERI and AECL.The fuel has demonstrated its irradiation performance in a Canadian commercial power reactor, Pt. Lepreau Generating Station since 1998 September. The RU(SEU) and DUPIC fuels are expected to be developed continuously until about the year 2010 for their use in CANDU reactors. Beside these fuel

  3. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO{sub 2}–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  4. Advances in the generation of a new emulsified fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, A. [Technical Consultancy, Energy Plus UC, Huitzilac, Morelos (Mexico); Ramirez, M. [Instituto Mexicano del Petroleo, Programa de Aseguramiento de Hidrocarburos, Mexico, D.F. (Mexico); Medina, E. [Universidad Nacional Autonoma de Mexico, Departamento de Termofluidos, Facultad de Ingenieria, Mexico, D.F. (Mexico); Bolado, R.; Mora, J. [Instituto Mexicano del Petroleo, Laboratorio de Combustion, Veracruz (Mexico)

    2011-08-15

    The development of a new emulsified fuel is described, from the conceptual idea to the semi-industrial tests of the final product. The starting point was the necessity to lower the particulate matter (PM) emissions produced by the combustion of more than 200 MBD of heavy fuel oil (HFO) used for electric power conversion. The major component of HFO is a vacuum residue of the oil refining process mixed with light cycle oils to make it pumpable. An alternative to handle and burn the high viscosity residue (solid at room temperature) is by converting it in an oil-in-water emulsion. The best emulsions resulted of 70% residue in 30% water, Sauter Mean Diameter of 10-20 {mu}m and a stability of more than 90 days. Spray burning tests of the emulsion against HFO in a semi-industrial 500 kW furnace showed a reduction in PM emissions of 24-36%. (orig.)

  5. Advanced ECU Software Development Method for Fuel Cell Systems

    Institute of Scientific and Technical Information of China (English)

    TIAN Shuo; LIU Yuan; XIA Wenchuan; LI Jianqiu; YANG Minggao

    2005-01-01

    The electronic control unit (ECU) in electrical powered hybrid and fuel cell vehicles is exceedingly complex. Rapid prototyping control is used to reduce development time and eliminate errors during software development. This paper describes a high-efficiency development method and a flexible tool chain suitable for various applications in automotive engineering. The control algorithm can be deployed directly from a Matlab/Simulink/Stateflow environment into the ECU hardware together with an OSEK real-time operating system (RTOS). The system has been successfully used to develop a 20-kW fuel cell system ECU based on a Motorola PowerPC 555 (MPC555) microcontroller. The total software development time is greatly reduced and the code quality and reliability are greatly enhanced.

  6. Recent advances in Carbon Nanotube based Enzymatic Fuel Cells

    Directory of Open Access Journals (Sweden)

    Serge eCosnier

    2014-10-01

    Full Text Available This review summarizes recent trends in the field of enzymatic fuel cells. Thanks to the high specificity of enzymes, biofuel cells can generate electrical energy by oxidation of a targeted fuel (sugars, alcohols or hydrogen at the anode and reduction of oxidants (O2, H2O2 at the cathode in complex media. The combination of carbon nanotubes, enzymes and redox mediators was widely exploited to develop biofuel cells since the electrons, involved in the bio-electrocatalytic processes, can be efficiently transferred from or to an external circuit. Original approaches to construct electron transfer based CNT-bioelectrodes and impressive biofuel cell performances are reported as well as biomedical applications.

  7. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  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. Functional nanocomposites for advanced fuel cell technology and polygeneration

    OpenAIRE

    Raza, Rizwan

    2011-01-01

    In recent decades, the use of fossil fuels has increased exponentially with a corresponding sharp increase in the pollution of the environment. The need for clean and sustainable technologies for the generation of power with reduced or zero environment impact has become critical. A number of attempts have been made to address this problem; one of the most promising attempts is polygeneration. Polygeneration technology is highly efficient and produces lower emissions than conventional methods ...

  10. Hybrid fuel cell bus demonstration: advanced technology moves bus forward

    International Nuclear Information System (INIS)

    The Province of Manitoba has been actively pursuing hydrogen since 2001 as one part of a portfolio of renewable energy alternatives. Six priority hydrogen actions have been underway covering a variety of opportunities, including two recently completed major transit bus and refueling demonstrations. A brief overview of Manitoba's activities on hydrogen will be provided, emphasizing the lessons learned from recent projects such as the hydrogen Hybrid Fuel Cell Bus demonstration, and in particular implications for the research community. (author)

  11. The proposed Fissile Material Cutoff Treaty (FMCT) and its potential impact on U.S. Navy nuclear propulsion programs

    OpenAIRE

    Burgess, Marion

    2010-01-01

    Approved for public release; distribution is unlimited This thesis examines the problems that United States Navy nuclear propulsion programs might encounter if the United States agreed to sign a version of the proposed Fissile Material Cut-off Treaty (FMCT) affecting the production of nuclear reactor fuel. The ultimate purpose of the FMCT is to contribute to the fulfillment of the goal of nuclear disarmament by terminating the production of plutonium and highly enriched uranium for we...

  12. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  13. DOE NCSP Review of TRUPACT-II/HalfPACT Fissile Limits

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2002-03-28

    The U.S. Department of Energy (DOE) Environmental Management (EM) Office of Nuclear Material & Spent Fuel, EM-21, tasked the CSSG to perform a scoping study to determine the feasibility of increasing the fissile mass loading limits for specified TRUPACT-II and HalfPACT packages and containers. The results of the scoping study may provide insights and technical guidance for establishing fissile mass loading limits at waste generator sites and at the waste repository. The goal is to reduce costs of transporting fissile material to the WIPP from EM's various closure sites. This report documents the results of the scoping study and demonstrates that it is feasible to significantly increase the fissile mass loading limits in the TRUPACT-II and HalfPACT packages and containers. Depending upon the particular payload containers used, the number of shipments to WIPP could be reduced by at least a factor of 2 and as much as a factor of 16 and the number of total payload containers required ''down-hole'' at WIPP could be reduced by at least a factor of 2 and as much as about 6. These cost savings result simply from applying a more realistic criticality analysis model rather than the very conservative, hypothetical, bounding analysis used to support the existing fissile mass loading limits. However, the applications of existing and developmental computational tools, nuclear data, and experiments from the DOE Nuclear Criticality Safety Program have the potential to further reduce transportation and disposal container costs on the order of 7% to 17%. It is suggested that EM proceed with an effort to do the required formal analyses and pursue SARP supplements to take advantage of these savings. The success of these analyses are dependent upon the availability of the majority of the infrastructure supported by the DOE Nuclear Criticality Safety Program as defined in the Five-Year Plan for the program. Finally, it should be noted that these potential cost

  14. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  15. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    International Nuclear Information System (INIS)

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the 'UREX+3c fuel cycle' and the 'Alternative Fuel Cycle' (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the

  16. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    Science.gov (United States)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  17. The nuclear fuel cycle with advanced reactor systems - analysis of its economic fundamentals and possibilities

    International Nuclear Information System (INIS)

    The purpose of the study is to analyse the nuclear fuel cycle of alternative advanced reactor systems with respect to their different mass flows of nuclear fuel and to judge the economic feasibility of these advanced nuclear technologies using a specific fuel cycle model. It is the particular importance of this subject that many technical, physical, political and economic coherences are combined in a very complex manner. A detailed description of the problem is given in the introductional chapter 1. The following chapter 2 gives a sufficient survey of the different techniques and technical facilities of the nuclear fuel cycles in question. Part 3 includes an investigation of logical coherences between typical fuel cycle mass flows which consequently leads to a mathematical model. This model is described in part 4. Chapter 5 then deals with the application of this model by the quantitative estimation and valuation of the economic differences between the conventional and advanced nuclear technology. In the final part of this study the influence of a very important parameter in this context, the price of plutonium, is discussed with respect to the time of introduction of the advanced reactor technology under economic conditions. (orig.)

  18. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  19. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  20. GOVERNOR ELECTRONICS FOR DIESEL ENGINES : High availability platform for real-time control and advanced fuel efficiency algorithms

    OpenAIRE

    Holmström, Johnny

    2013-01-01

    Fossil fuel is a rare commodity and the combustion of this fuel results in negative environmental effects. This paper evaluates and validates the electronics needed to run intelligent algorithms to lower the fuel consumption for commercial vessels. This is done by integrating advanced fuel saving functions into an electronic device that controls the fuel injection of large diesel engines, as known as a diesel engine governor. The control system is classified as a safety critical system. This ...

  1. Study on process basic requirements of experimental facility of advanced spent fuel management process

    International Nuclear Information System (INIS)

    The advanced spent fuel management process, which was proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. Hot cell facilities of α-γ type and inert atmosphere are required essentially for safe hot test and verification of this process. In this study, design basic data are established, and these data include process flow, process condition and yields, mass and radioactivity balance of radionuclides, process safety considerations, etc. And also, these data will be utilized for basic and detail design of hot cell facility, secured conservative safety and effective operability

  2. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2013-01-01

    SunLine Transit Agency, which provides public transit services to the Coachella Valley area of California, has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technology (AT) fuel cell bus with a hybrid electric propulsion system, fuel cell power system, and lithium-based hybrid batteries. This report describes operations at SunLine for the AT fuel cell bus and five compressed natural gas buses. The U.S. Department of Energy's National Renewable Energy Laboratory (NREL) is working with SunLine to evaluate the bus in real-world service to document the results and help determine the progress toward technology readiness. NREL has previously published three reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from February 2012 through November 2012.

  3. Feasible advanced fuel cycle options for CANDU reactors in the Republic of Korea

    International Nuclear Information System (INIS)

    Taking into account the view points on nuclear safety, nuclear waste, non-proliferation and economics from the public, international environment, and utilities, the SEU/RU and DUPIC fuel cycles would be feasible options of advanced fuel cycles for CANDU-PHWRs in the Republic of Korea in the mid- and long-terms, respectively. Comparing with NU fuel, 0.9 % or 1.2 % SEU fuel would increase fuel burnup and hence reduce the spent fuel arisings by a factor of 2 or 3, and also could reduce CANDU fuel cycle costs by 20 to 30%. RU offers similar benefits as 0.9% SEU and is very attractive due to the significantly improved fuel cycle economics, substantially increased burnups, large reduction in fuel requirements as well as in spent fuel arisings. For RU use in a CANDU reactor, re-enrichment is not required. There are 25,000 tes RU produced from reprocessing operations in Europe and Japan, which would theoretically provide sufficient fuel for 500 CANDU 6 reactor-years of operation. According to the physics, thermal-hydraulic and thermal-mechanical assessments of CANFLEX-0.9% RU fuel for a CANDU-6 reactor, the fuel could be introduced into the reactor in a straight-forward fashion. A series of assessments of CANFLEX-DUPIC physics on the compatibility of the fuel design in the existing CANDU 6 reactors has shown that the poisoning of the central element of DUPIC with, for example, natural dysprosium, reduces the void reactivity of the fuel, and that a 2 bundle shift refuelling scheme would be the most appropriate in-core fuel management scheme for a CANDU-6 reactor. The average discharge burnup is ∼15 MWd/kgHE. Although these results have shown promising results for the DUPIC fuel cycle, more in-depth studies are required in the areas of ROP system, large LOCA safety analyses, and so on. The recycling fuel cycles of RU and DUPIC for CANDU are expected to achieve the environmental 3R's (Reduce, Reuse, Recycle) as applied to global energy use in the short- and long

  4. Advanced Automotive Fuels Research, Development, and Commercialization Cluster (OH)

    Energy Technology Data Exchange (ETDEWEB)

    Linkous, Clovis; Hripko, Michael; Abraham, Martin; Balendiran, Ganesaratnam; Hunter, Allen; Lovelace-Cameron, Sherri; Mette, Howard; Price, Douglas; Walker, Gary; Wang, Ruigang

    2013-08-31

    Technical aspects of producing alternative fuels that may eventually supplement or replace conventional the petroleum-derived fuels that are presently used in vehicular transportation have been investigated. The work was centered around three projects: 1) deriving butanol as a fuel additive from bacterial action on sugars produced from decomposition of aqueous suspensions of wood cellulose under elevated temperature and pressure; 2) using highly ordered, openly structured molecules known as metal-organic framework (MOF) compounds as adsorbents for gas separations in fuel processing operations; and 3) developing a photocatalytic membrane for solar-driven water decomposition to generate pure hydrogen fuel. Several departments within the STEM College at YSU contributed to the effort: Chemistry, Biology, and Chemical Engineering. In the butanol project, sawdust was blended with water at variable pH and temperature (150 – 250{degrees}C), and heated inside a pressure vessel for specified periods of time. Analysis of the extracts showed a wide variety of compounds, including simple sugars that bacteria are known to thrive upon. Samples of the cellulose hydrolysate were fed to colonies of Clostridium beijerinckii, which are known to convert sugars to a mixture of compounds, principally butanol. While the bacteria were active toward additions of pure sugar solutions, the cellulose extract appeared to inhibit butanol production, and furthermore encouraged the Clostridium to become dormant. Proteomic analysis showed that the bacteria had changed their genetic code to where it was becoming sporulated, i.e., the bacteria were trying to go dormant. This finding may be an opportunity, as it may be possible to genetically engineer bacteria that resist the butanol-driven triggering mechanism to stop further fuel production. Another way of handling the cellulosic hydrolysates was to simply add the enzymes responsible for butanol synthesis to the hydrolytic extract ex-vivo. These

  5. The development of the integral version of the neutron resonance transmission analysis to solve the problem of estimating fissile elements in spent nuclear assemblies

    International Nuclear Information System (INIS)

    The paper describes the development of the integral version of the neutron resonance transmission analysis (NRTA) to solve the problem of estimating fissile elements in spent nuclear fuel. Using mathematical modeling involving the ENDF/B5 nuclear data library, the range of material thickness has been identified for which the technique is the most effective one

  6. 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

  7. THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, Bartley B.; Collins, Brian A.; Sleaford, Brad W.; Hase, Kevin R.; Robel, Martin; Wallace, R. K.; Bradley, Keith S.; Ireland, J. R.; Jarvinen, G. D.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2012-08-29

    We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

  8. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    An inexpensive boron-loaded liner of epoxy resin for fissile-material storage containers was developed that can be easily fabricated of readily available, low-cost materials. Computer calculations indicate reactivity will be reduced substantially if this neutron-absorbing liner is added to containers in a typical storage array. These calculations compare favorably with neutron-attenuation experiments with thermal and fission neutron spectra, and tests at the Fire Test Facility indicate the epoxy resin will survive extreme environmental and accident conditions. The fire-resistant and insulating properties of the epoxy-resin liner further augment its ability to protect fissile materials. Boron-loaded epoxy resin is adaptable to many tasks but is particularly useful for providing enhanced criticality safety in the packaging and storage of fissile materials

  9. Fissile and fertile nuclear material measurements using a new differential die-away self-interrogation technique

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard O [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Menlove, S H [Los Alamos National Laboratory

    2008-01-01

    This paper presents a new technique for the measurement of fissile and fertile nuclear materials in spent fuel and plutonium laden materials such as mixed oxide (MOX) fuel. The technique, called differential die-away self-interrogation, is similar to traditional differential die-away analysis, but it does not require a pulsed neutron generator or pulsed beam accelerator, and it can measure the fertile mass in addition to the fissile mass. The new method uses the spontaneous fission neutrons from {sup 244}Cm in spent fuel and {sup 240}Pu effective neutrons in MOX as the 'pulsed' neutron source with an average of {approx} 2.7 neutrons per pulse. The time correlated neutrons from the spontaneous fission and the subsequent induced fissions are analyzed as a function of time to determine the spontaneous fission rate, the induced fast-neutron fissions, and the induced thermal-neutron fissions. The fissile mass is determined from the induced thermal-neutron fissions that are produced by reflected thermal neutrons that originated from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by the use of two measurements, with and without a cadmium liner between the sample and the hydrogenous moderator. The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. The method obtains good sensitivity by the optimal design of two different neutron die-away regions: a short die-away for the neutron detector region and a longer die-away for the sample interrogation region.

  10. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Duane D. Bruns; Robert M. Counce; Irma D. Lima Rojas

    2010-06-09

    this research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.

  11. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath; Donald G. Baird; Michael von Spakovsky

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 degrees C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and states of water in the membranes. Further improvements in properties were achieved through incorporation of inorganic

  12. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 °C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and

  13. Development of advanced nuclear fuels in the Indian context: advantages and challenges

    International Nuclear Information System (INIS)

    The ever increasing demand on power requirement in the country has opened up need for exploring use of nuclear fuels that could meet such demands. This makes the mission of the department to shift from the first stage of nuclear programme employing natural uranium in PHWRs to the second stage of deploying a large number of fast reactors with plutonium based fuels capable of realising high breeding ratios in addition to energy production. The transition to fast reactors with advanced fuels, capable of higher breeding ratio, opens up a number of scientific and technological challenges in design and operation of such fast reactors. In the Indian context, after successful demonstration of natural uranium based PHWRs, the performance of U-Pu based carbide fuel, as a unique experience in the world, has been demonstrated in FBTR at Kalpakkam. This paper deals with the performance of carbide fuel in FBTR and the programme on development of metallic fuels with appreciably high breeding ratio that would result in considerable reduction in doubling time thereby addressing the increasing demands of power production as well as pave way for introduction of a large number of such fast reactors to provide energy security to the country. The advantages of introduction of metallic fuels as well as the scientific and technological challenges to be faced in doing so and the ongoing efforts towards metallic fuel development are also described in the paper. (author)

  14. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    Science.gov (United States)

    Delage, F.; Carmack, J.; Lee, C. B.; Mizuno, T.; Pelletier, M.; Somers, J.

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

  15. A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future

    Energy Technology Data Exchange (ETDEWEB)

    Maiorino, J.R.; Carluccio, T.

    2004-10-03

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.

  16. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated 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 was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  17. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  18. Stress analysis for the candidate of lower end fitting of advanced LWR fuel using FEM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. S.; Moon, Y. C. [Korea University of Technology and Education, Chonan (Korea, Republic of); Kim, H. K. [Korea Nuclear Fuel Company, Taejon (Korea, Republic of)

    2002-10-01

    The geometric modeling has been conducted for the candidate of advanced LWR fuel using the three-dimensional solid modeler. Then the three-dimensional stress analysis using MSC/NASTRAN has been performed. The evaluation for the mechanical integrity of the candidate has been performed based on the stress distribution obtained from the finite elements analysis.

  19. Leo Szilard Lectureship Award: Fissile Materials: A Global Threat

    Science.gov (United States)

    Rajaraman, Ramamurti

    2014-03-01

    The world has built up a huge glut of Fissile Materials, posing a potentially devastating threat. While specialists in the field have been aware of this danger for a long time, it was only after President Obama organized the Nuclear Security Summit in 2010 that the attention of the world's political leadership was drawn to it. We will present here an introductory overview of Fissile materials - their definition, significance and their production facilities and stocks in different parts of the world. We will also mention some of the efforts being made to verifiably cap and reduce their stocks as well as the technical and political complications involved in the process.

  20. Impact of fuel properties on advanced power systems

    Energy Technology Data Exchange (ETDEWEB)

    Sondreal, E.A.; Jones, M.L.; Hurley, J.P.; Benson, S.A.; Willson, W.G. [Univ. of North Dakota, Grand Forks, ND (United States)

    1995-12-01

    Advanced coal-fired combined-cycle power systems currently in development and demonstration have the goal of increasing generating efficiency to a level approaching 50% while reducing the cost of electricity from new plants by 20% and meeting stringent standards on emissions of SO{sub x} NO{sub x} fine particulates, and air toxic metals. Achieving these benefits requires that clean hot gas be delivered to a gas turbine at a temperature approaching 1350{degrees}C, while minimizing energy losses in the gasification, combustion, heat transfer, and/or gas cleaning equipment used to generate the hot gas. Minimizing capital cost also requires that the different stages of the system be integrated as simply and compactly as possible. Second-generation technologies including integrated gasification combined cycle (IGCC), pressurized fluidized-bed combustion (PFBC), externally fired combined cycle (EFCC), and other advanced combustion systems rely on different high-temperature combinations of heat exchange, gas filtration, and sulfur capture to meet these requirements. This paper describes the various properties of lignite and brown coals.

  1. Annual Report: Advanced Energy Systems Fuel Cells (30 September 2013)

    Energy Technology Data Exchange (ETDEWEB)

    Gerdes, Kirk; Richards, George

    2014-04-16

    The comprehensive research plan for Fuel Cells focused on Solid State Energy Conversion Alliance (SECA) programmatic targets and included objectives in two primary and focused areas: (1) investigation of degradation modes exhibited by the anode/electrolyte/cathode (AEC), development of computational models describing the associated degradation rates, and generation of a modeling tool predicting long term AEC degradation response; and (2) generation of novel electrode materials and microstructures and implementation of the improved electrode technology to enhance performance. In these areas, the National Energy Technology Laboratory (NETL) Regional University Alliance (RUA) team has completed and reported research that is significant to the SECA program, and SECA continued to engage all SECA core and SECA industry teams. Examination of degradation in an operational solid oxide fuel cell (SOFC) requires a logical organization of research effort into activities such as fundamental data gathering, tool development, theoretical framework construction, computational modeling, and experimental data collection and validation. Discrete research activity in each of these categories was completed throughout the year and documented in quarterly reports, and researchers established a framework to assemble component research activities into a single operational modeling tool. The modeling framework describes a scheme for categorizing the component processes affecting the temporal evolution of cell performance, and provides a taxonomical structure of known degradation processes. The framework is an organizational tool that can be populated by existing studies, new research completed in conjunction with SECA, or independently obtained. The Fuel Cell Team also leveraged multiple tools to create cell performance and degradation predictions that illustrate the combined utility of the discrete modeling activity. Researchers first generated 800 continuous hours of SOFC experimental

  2. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    facilities. - 3. Advances in Water Reactor Fuel Technology: Advances in fuel, rod, spacer grids, and assembly design; fuel processing and manufacturing; cladding and structural alloy development; MOX fuel design and manufacturing; advances in fuel pellet development; fuel design for improved thermal hydraulics, mechanical, and corrosion-resistant behavior; irradiation experience in test reactors. - 4. Concepts for Transportation and Interim Storage of Spent Fuels and Conditioned Waste (Shared with Global 2009): Industrial experience and ongoing developments. - 5. Innovative Fuel Design and Core Management: Future development and trends in fuel for the next thirty years; Goals and perspectives for nuclear fuel; Long term improvement in fissile material management; Use of composite material; Innovative microstructure and material under development; Future core management.

  3. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    Reactor Fuel Technology: Advances in fuel, rod, spacer grids, and assembly design; fuel processing and manufacturing; cladding and structural alloy development; MOX fuel design and manufacturing; advances in fuel pellet development; fuel design for improved thermal hydraulics, mechanical, and corrosion-resistant behavior; irradiation experience in test reactors. - 4. Concepts for Transportation and Interim Storage of Spent Fuels and Conditioned Waste (Shared with Global 2009): Industrial experience and ongoing developments. - 5. Innovative Fuel Design and Core Management: Future development and trends in fuel for the next thirty years; Goals and perspectives for nuclear fuel; Long term improvement in fissile material management; Use of composite material; Innovative microstructure and material under development; Future core management

  4. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  5. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUELS PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    Shane E. Roark; Anthony F. Sammells; Richard Mackay; Stewart Schesnack; Scott Morrison; Thomas A. Zirbel; Thomas F. Barton; Sara L. Rolfe; U. Balachandran; Richard N. Kleiner; James E. Stephan; Frank E. Anderson; Aaron L. Wagner; Jon P. Wagner

    2003-07-31

    Eltron Research Inc. and team members CoorsTek, Sued Chemie, and Argonne National Laboratory are developing an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. This project was motivated by the National Energy Technology Laboratory (NETL) Vision 21 initiative, which seeks to economically eliminate environmental concerns associated with the use of fossil fuels. Currently, this project is focusing on four basic categories of dense membranes: (1) mixed conducting ceramic/ceramic composites, (2) mixed conducting ceramic/metal (cermet) composites, (3) cermets with hydrogen permeable metals, and (4) layered composites containing hydrogen permeable alloys. Ultimately, these materials must enable hydrogen separation at practical rates under ambient and high-pressure conditions, without deactivation in the presence of feedstream components such as carbon dioxide, water, and sulfur. This report presents hydrogen permeation data during long term tests and tests at high pressure in addition to progress with cermet, ceramic/ceramic, and thin film membranes.

  6. Economic and Environmental Value of Advanced Fuel Cycle

    International Nuclear Information System (INIS)

    The goal of AFC is to achieve a significant reduction of High Level Waste (HLW) and accumulated plutonium in the SNF through Partitioning and Transmutation (P and T), and to recover the useful materials from the SNF. Because of its technological advantages in many aspects, its possibility of realization was tested and supported by many studies and works. The economic value of AFC has been the main concern since its development, albeit the bigger merit in other aspects. In this study, therefore, another value, namely the environmental value, will be discussed and the sum will also be considered. In the environmental value, significant merits over direct disposal were achieved by reduced accumulation of the SNFs and less purchased uranium for reactor fuel. It can be concluded that the total value of the AFC can be greater than that of direct disposal, if the required condition is set. For further extension of this study, consideration of safeguard and social value for each cycle will provide important information

  7. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  8. Advanced multi-fuelled solid oxide fuel cells (ASOFCs) using functional nanocomposites for polygeneration

    Energy Technology Data Exchange (ETDEWEB)

    Raza, Rizwan [Department of Physics, COMSATS Institute of Information Technology, Lahore (Pakistan); Department of Energy Technology, Royal Institute of Technology, Stockholm (Sweden); Qin, Haiying; Samavati, Mahrokh; Zhu, Bin [Department of Energy Technology, Royal Institute of Technology, Stockholm (Sweden); Liu, Qinghua [Tianjin Laboratory for Chemical Engineering (Tianjin University), School of Chemical Engineering and Technology, Tianjin University, Tianjin, 300072 (China); Lima, Raquel B. [Department of Fiber and Polymer Technology, Royal Institute of Technology, KTH, 10044, Stockholm (Sweden)

    2011-11-15

    An advanced multifuelled solid oxide fuel cell (ASOFC) with a functional nanocomposite was developed and tested for use in a polygeneration system. Several different types of fuel, for example, gaseous (hydrogen and biogas) and liquid fuels (bio-ethanol and bio-methanol), were used in the experiments. Maximum power densities of 1000, 300, 600, 550 mW cm{sup -2} were achieved using hydrogen, bio-gas, bio-methanol, and bio-ethanol, respectively, in the ASOFC. Electrical and total efficiencies of 54% and 80% were achieved using the single cell with hydrogen fuel. These results show that the use of a multi-fuelled system for polygeneration is a promising means of generating sustainable power. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  9. Clean Cities Guide to Alternative Fuel and Advanced Medium- and Heavy-Duty Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-08-01

    Today's fleets are increasingly interested in medium-duty and heavy-duty vehicles that use alternative fuels or advanced technologies that can help reduce operating costs, meet emissions requirements, improve fleet sustainability, and support U.S. energy independence. Vehicle and engine manufacturers are responding to this interest with a wide range of options across a steadily growing number of vehicle applications. This guide provides an overview of alternative fuel power systems--including engines, microturbines, electric motors, and fuel cells--and hybrid propulsion systems. The guide also offers a list of individual medium- and heavy-duty vehicle models listed by application, along with associated manufacturer contact information, fuel type(s), power source(s), and related information.

  10. Clean Cities Guide to Alternative Fuel and Advanced Medium- and Heavy-Duty Vehicles (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2013-08-01

    Today's fleets are increasingly interested in medium-duty and heavy-duty vehicles that use alternative fuels or advanced technologies that can help reduce operating costs, meet emissions requirements, improve fleet sustainability, and support U.S. energy independence. Vehicle and engine manufacturers are responding to this interest with a wide range of options across a steadily growing number of vehicle applications. This guide provides an overview of alternative fuel power systems?including engines, microturbines, electric motors, and fuel cells?and hybrid propulsion systems. The guide also offers a list of individual medium- and heavy-duty vehicle models listed by application, along with associated manufacturer contact information, fuel type(s), power source(s), and related information.

  11. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Third Results Reports

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2012-05-01

    This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. NREL has previously published two reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from July 2011 through January 2012.

  12. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Second Results Report and Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2011-10-01

    This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This is the second results report for the AT fuel cell bus since it was placed in service, and it focuses on the newest data analysis and lessons learned since the previous report. The appendices, referenced in the main report, provide the full background for the evaluation. They will be updated as new information is collected but will contain the original background material from the first report.

  13. Design concepts and advanced manipulator development for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    In the Fuel Recycle Division, Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory, a comprehensive remote systems development program has existed for the past seven years. The new remote technology under development is expected to significantly improve remote operations by extending the range of tasks accomplished by remote means and increasing the efficiency of remote work undertaken. The application of advanced manipulation is viewed as an essential part of a series of design directions whose sum describes a somewhat unique blend of old and new technology. A design direction based upon the Teletec concept is explained and recent progress in the development of an advanced servomanipulator-based maintenance concept is summarized to show that a new generation of remote systems is feasible through advanced technology. 14 refs., 14 figs

  14. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  15. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

  16. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  17. Fuel Distribution Estimate via Spin Period to Precession Period Ratio for the Advanced Composition Explorer

    Science.gov (United States)

    DeHart, Russell; Smith, Eric; Lakin, John

    2015-01-01

    The spin period to precession period ratio of a non-axisymmetric spin-stabilized spacecraft, the Advanced Composition Explorer (ACE), was used to estimate the remaining mass and distribution of fuel within its propulsion system. This analysis was undertaken once telemetry suggested that two of the four fuel tanks had no propellant remaining, contrary to pre-launch expectations of the propulsion system performance. Numerical integration of possible fuel distributions was used to calculate moments of inertia for the spinning spacecraft. A Fast Fourier Transform (FFT) of output from a dynamics simulation was employed to relate calculated moments of inertia to spin and precession periods. The resulting modeled ratios were compared to the actual spin period to precession period ratio derived from the effect of post-maneuver nutation angle on sun sensor measurements. A Monte Carlo search was performed to tune free parameters using the observed spin period to precession period ratio over the life of the mission. This novel analysis of spin and precession periods indicates that at the time of launch, propellant was distributed unevenly between the two pairs of fuel tanks, with one pair having approximately 20% more propellant than the other pair. Furthermore, it indicates the pair of the tanks with less fuel expelled all of its propellant by 2014 and that approximately 46 kg of propellant remains in the other two tanks, an amount that closely matches the operational fuel accounting estimate. Keywords: Fuel Distribution, Moments of Inertia, Precession, Spin, Nutation

  18. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  19. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  20. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  1. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    International Nuclear Information System (INIS)

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  2. Evolutionary/advanced light water reactor data report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  3. ADVANCING THE FUNDAMENTAL UNDERSTANDING AND SCALE-UP OF TRISO FUEL COATERS VIA ADVANCED MEASUREMENT AND COMPUTATIONAL TECHNIQUES

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, Pratim; Al-Dahhan, Muthanna

    2012-11-01

    Tri-isotropic (TRISO) fuel particle coating is critical for the future use of nuclear energy produced byadvanced gas reactors (AGRs). The fuel kernels are coated using chemical vapor deposition in a spouted fluidized bed. The challenges encountered in operating TRISO fuel coaters are due to the fact that in modern AGRs, such as High Temperature Gas Reactors (HTGRs), the acceptable level of defective/failed coated particles is essentially zero. This specification requires processes that produce coated spherical particles with even coatings having extremely low defect fractions. Unfortunately, the scale-up and design of the current processes and coaters have been based on empirical approaches and are operated as black boxes. Hence, a voluminous amount of experimental development and trial and error work has been conducted. It has been clearly demonstrated that the quality of the coating applied to the fuel kernels is impacted by the hydrodynamics, solids flow field, and flow regime characteristics of the spouted bed coaters, which themselves are influenced by design parameters and operating variables. Further complicating the outlook for future fuel-coating technology and nuclear energy production is the fact that a variety of new concepts will involve fuel kernels of different sizes and with compositions of different densities. Therefore, without a fundamental understanding the underlying phenomena of the spouted bed TRISO coater, a significant amount of effort is required for production of each type of particle with a significant risk of not meeting the specifications. This difficulty will significantly and negatively impact the applications of AGRs for power generation and cause further challenges to them as an alternative source of commercial energy production. Accordingly, the proposed work seeks to overcome such hurdles and advance the scale-up, design, and performance of TRISO fuel particle spouted bed coaters. The overall objectives of the proposed work are

  4. Fissile mass estimation by pulsed neutron source interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Israelashvili, I., E-mail: israelashvili@gmail.com [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Dubi, C.; Ettedgui, H.; Ocherashvili, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Pedersen, B. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Beck, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Roesgen, E.; Crochmore, J.M. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Ridnik, T.; Yaar, I. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel)

    2015-06-11

    Passive methods for detecting correlated neutrons from spontaneous fissions (e.g. multiplicity and SVM) are widely used for fissile mass estimations. These methods can be used for fissile materials that emit a significant amount of fission neutrons (like plutonium). Active interrogation, in which fissions are induced in the tested material by an external continuous source or by a pulsed neutron source, has the potential advantages of fast measurement, alongside independence of the spontaneous fissions of the tested fissile material, thus enabling uranium measurement. Until recently, using the multiplicity method, for uranium mass estimation, was possible only for active interrogation made with continues neutron source. Pulsed active neutron interrogation measurements were analyzed with techniques, e.g. differential die away analysis (DDA), which ignore or implicitly include the multiplicity effect (self-induced fission chains). Recently, both, the multiplicity and the SVM techniques, were theoretically extended for analyzing active fissile mass measurements, made by a pulsed neutron source. In this study the SVM technique for pulsed neutron source is experimentally examined, for the first time. The measurements were conducted at the PUNITA facility of the Joint Research Centre in Ispra, Italy. First promising results, of mass estimation by the SVM technique using a pulsed neutron source, are presented.

  5. Safeguards and security issues for the disposition of fissile materials

    International Nuclear Information System (INIS)

    The Department of Energy's Office of Fissile Material Disposition (FMD) is analyzing long-term storage and disposition options for surplus weapons-usable fissile materials, preparing a programmatic environmental impact statement (PEIS), preparing for a record of decision (ROD) regarding this material and conducting other activities. The primary security objectives of this program are to reduce major security risks and strengthen arms reduction and nonproliferation (NP). To help achieve these objectives, a safeguards and security (S ampersand S) team consisting of participants from Sandia, Los Alamos, and Lawrence Livermore National Laboratories was established. The S ampersand S activity for this program is a cross-cutting task which addresses all of the FMD program options. It includes both domestic and international safeguards and includes areas such as physical protection, nuclear materials accountability and material containment and surveillance. This paper will discuss the activities of the Fissile Materials Disposition Program (FMDP) S ampersand S team as well as some specific S ampersand S issues associated with various FMDP options/facilities. Some of the items to be discussed include the threat, S ampersand S requirements, S ampersand S criteria for assessing risk, S ampersand S issues concerning fissile material processing/facilities, and international and domestic safeguards

  6. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    Shane E. Roark; Anthony F. Sammells; Richard Mackay; Stewart R. Schesnack; Scott R. Morrison; Thomas F. Barton; Sara L. Rolfe; U. Balachandran; Richard N. Kleiner; James E. Stephan; Frank E. Anderson; Aaron L. Wagner; Jon P. Wagner

    2003-10-30

    Eltron Research Inc. and team members CoorsTek, Sued Chemie, Argonne National Laboratory, and NORAM are developing an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. This project was motivated by the National Energy Technology Laboratory (NETL) Vision 21 initiative, which seeks to economically eliminate environmental concerns associated with the use of fossil fuels. Over the past 12 months, this project has focused on four basic categories of dense membranes: (1) mixed conducting ceramic/ceramic composites, (2) mixed conducting ceramic/metal (cermet) composites, (3) cermets with hydrogen permeable metals, and (4) layered composites containing hydrogen permeable alloys. Ultimately, these materials must enable hydrogen separation at practical rates under ambient and high-pressure conditions, without deactivation in the presence of feedstream components such as carbon dioxide, water, and sulfur. The ceramic/ceramic composites demonstrate the lowest hydrogen permeation rates, with a maximum of approximately 0.1 mL/min/cm{sup 2} for 0.5-mm thick membranes at 800 to 950 C. Under equivalent conditions, cermets achieve a hydrogen permeation rate near 1 mL/min/cm{sup 2}, and the metal phase also improves structural stability and surface catalysis for hydrogen dissociation. Furthermore, if metals with high hydrogen permeability are used in cermets, permeation rates near 4 mL/min/cm{sup 2} are achievable with relatively thick membranes. Layered composite membranes have by far the highest permeation rates with a maximum flux in excess of 200 mL {center_dot} min{sup -1} {center_dot} cm{sup -2}. Moreover, these permeation rates were achieved at a total pressure differential across the membrane of 450 psi. Based on these results, effort during the next year will focus on this category of membranes. This report contains long-term hydrogen permeation data over eight

  7. Intergovernmental Advanced Stationary PEM Fuel Cell System Demonstration Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rich Chartrand

    2011-08-31

    efficiency and reducing costs of PEMFC based power systems using LPG fuel and continues to makes steps towards meeting DOE's targets. Plug Power would like to thank DOE for their support of this program.

  8. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

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

  9. 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

  10. Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

  11. The JRC-ITU approach to the safety of advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimental facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU

  12. Follow-up fuel plate stability experiments and analyses for the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

    1993-11-01

    The reactor for the planned Advanced Neutron Source uses closely spaced plates cooled by heavy water flowing through narrow channels. Two sets of tests were performed on the upper and lower fuel plates for the structural response of the fuel plates to the required high coolant flow velocities. This report contains the data from the second round of tests. Results and conclusions from all of the tests are also included in this report. The tests were done using light water on full-scale epoxy models, and through model theory, the results were related to the prototype plates, which are aluminum-clad aluminum/uranium silicide involute-shaped plates.

  13. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    International Nuclear Information System (INIS)

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  14. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  15. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    technological advances attempted in doping of fuel pellets with the primary objective of obtaining larger grains. While most of the papers gave an account of the experimental studies on addition of various dopants in different fuel materials, some of them outlined the behaviour of such pellets at sintering process. Papers dealing with 'Fission gas release from fuel pellets under high burnup conditions were presented in Session 3. Session 4 was devoted to the evolution of fuel pellet structure and thermal properties at high burnup. Session 5 was dealing with fuel pellet-cladding interaction (PCI) being a complex phenomenon that may lead to cladding failure and subsequent release of fission products into the reactor coolant. Research efforts to understand better the PCI phenomenon and minimize it with design solutions are considered necessary

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

    International Nuclear Information System (INIS)

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

  17. Uncertainty analysis of delayed neutron fissile material assay using a genetic algorithm

    International Nuclear Information System (INIS)

    Highlights: • This work assessed the usefulness of the delayed neutron signal for quantifying isotopic masses of three fissile materials in a sample. • A genetic algorithm was used to find the combination of fissile masses that best fit the delayed neutron signal. • For the first time, the uncertainties associated with this method were published. • The sensitivity of the algorithm to various nuclear data, particularly the relative delayed neutron group abundances, was also investigated. - Abstract: An uncertainty analysis of non-destructive assay of spent fuel using a delayed neutron method was conducted to explicitly define the accuracy with which plutonium content in an uncharacterized sample can be assessed. Perturbing various parameters allowed for an investigation of the sensitivity of this method to various nuclear data, and it was determined that the relative delayed neutron group abundances had the largest effect on the genetic algorithm. Specifically, for a sample containing 235U, 238U, and 239Pu, irradiations in the thermal spectrum were shown to be more sensitive to 235U and 239U data, while irradiations in a fast spectrum were shown to be more sensitive to the 238U data. The overall uncertainties of the mass estimates were 15%, 5%, and 30% for 235U, 238U, and 239Pu, respectively. Finally, reducing the first delayed neutron group abundances by a factor of three as suggested by recent research reduced the overall uncertainties to 10%, 3%, and 20%

  18. 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.

  19. Advanced coal-fueled gas turbine systems: Subscale combustion testing. Topical report, Task 3.1

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This is the final report on the Subscale Combustor Testing performed at Textron Defense Systems` (TDS) Haverhill Combustion Laboratories for the Advanced Coal-Fueled Gas Turbine System Program of the Westinghouse Electric Corp. This program was initiated by the Department of Energy in 1986 as an R&D effort to establish the technology base for the commercial application of direct coal-fired gas turbines. The combustion system under consideration incorporates a modular staged, rich-lean-quench, Toroidal Vortex Slogging Combustor (TVC) concept. Fuel-rich conditions in the first stage inhibit NO{sub x} formation from fuel-bound nitrogen; molten coal ash and sulfated sorbent are removed, tapped and quenched from the combustion gases by inertial separation in the second stage. Final oxidation of the fuel-rich gases, and dilution to achieve the desired turbine inlet conditions are accomplished in the third stage, which is maintained sufficiently lean so that here, too, NO{sub x} formation is inhibited. The primary objective of this work was to verify the feasibility of a direct coal-fueled combustion system for combustion turbine applications. This has been accomplished by the design, fabrication, testing and operation of a subscale development-type coal-fired combustor. Because this was a complete departure from present-day turbine combustors and fuels, it was considered necessary to make a thorough evaluation of this design, and its operation in subscale, before applying it in commercial combustion turbine power systems.

  20. Applications study of advanced power generation systems utilizing coal-derived fuels. Volume 1: Executive summary

    Science.gov (United States)

    Robson, F. L.

    1981-03-01

    The technology status of phosphoric acid and molten carbon fuel cells, combined gas and steam turbine cycles, and magnetohydrodynamic energy conversion systems was assessed and the power performance of these systems when operating with medium-Btu fuel gas whether delivered by pipeline to the power plant or in an integrated mode in which the coal gasification process and power system are closely coupled as an overall power plant was evaluated. Commercially available combined-cycle gas turbine systems can reach projected required performance levels for advanced systems using currently available technology. The phosphoric acid fuel cell appears to be the next most likely candidate for commercialization. On pipeline delivery, the systems efficiency ranges from 40.9% for the phosphoric acid fuel cell to 63% for the molten carbonate fuel cell system. The efficiencies of the integrated power plants vary from approximately 39-40% for the combined cycle to 46-47% for the molden carbonate fuel cell systems. Conventional coal-fired steam stations with flue-gas desulfurization have only 33-35% efficiency.

  1. 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.

  2. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  3. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

  4. Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

  5. Recent advances in fuel fabrication techniques and prospects for the nineties

    International Nuclear Information System (INIS)

    Advanced Nuclear Fuels Corporation's approach and experience with the application of a flexible, just-in-time manufacturing philosophy to the production of customized nuclear fuel is described. Automation approaches to improve productivity are described. The transfer of technology across product lines is discussed as well as the challenges presented by a multiple product fabrication facility which produces a wide variety of BWR and PWR designs. This paper also describes the method of managing vendor quality control programs in support of standardization and clarity of documentation. Process simplification and the ensuing experience are discussed. Prospects for fabrication process advancements in the nineties are given with emphasis on the benefits of dry conversion of UF6 to UO2 powder, and increased use of automated and computerized inspection techniques. (author)

  6. Performance and economics of advanced energy conversion systems for coal and coal-derived fuels

    Science.gov (United States)

    Corman, J. C.; Fox, G. R.

    1978-01-01

    The desire to establish an efficient Energy Conversion System to utilize the fossil fuel of the future - coal - has produced many candidate systems. A comparative technical/economic evaluation was performed on the seven most attractive advanced energy conversion systems. The evaluation maintains a cycle-to-cycle consistency in both performance and economic projections. The technical information base can be employed to make program decisions regarding the most attractive concept. A reference steam power plant was analyzed to the same detail and, under the same ground rules, was used as a comparison base. The power plants were all designed to utilize coal or coal-derived fuels and were targeted to meet an environmental standard. The systems evaluated were two advanced steam systems, a potassium topping cycle, a closed cycle helium system, two open cycle gas turbine combined cycles, and an open cycle MHD system.

  7. Advanced coal-fueled industrial cogeneration gas turbine system. Annual report, June 1990--June 1991

    Energy Technology Data Exchange (ETDEWEB)

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; Wen, C.S.

    1991-07-01

    Advances in coal-fueled gas turbine technology over the past few years, together with recent DOE-METC sponsored studies, have served to provide new optimism that the problems demonstrated in the past can be economically resolved and that the coal-fueled gas turbine can ultimately be the preferred system in appropriate market application sectors. The objective of the Solar/METC program is to prove the technical, economic, and environmental feasibility of a coal-fired gas turbine for cogeneration applications through tests of a Centaur Type H engine system operated on coal fuel throughout the engine design operating range. The five-year program consists of three phases, namely: (1) system description; (2) component development; (3) prototype system verification. A successful conclusion to the program will initiate a continuation of the commercialization plan through extended field demonstration runs.

  8. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  9. Recent advances during the treatment of spent EBR-II fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B. R.; Mariani, R. D.; Vaden, D. E.; Sherman, S. R.; Li, S. X.; Keiser, D. D., Jr.

    2000-03-20

    Several recent advances have been achieved for the electrometallurgical treatment of spent nuclear fuel. In anticipation of production operations at Argonne National Laboratory-West, development of both electrorefining and metal processing has been ongoing in the post-demonstration phase in order to further optimize the process. These development activities show considerable promise. This paper discusses the results of recent experiments as well as plans for future investigations.

  10. Advanced combustion, emission control, health impacts, and fuels merit review and peer evaluation

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2006-10-01

    This report is a summary and analysis of comments from the Advisory Panel at the FY 2006 DOE National Laboratory Advanced Combustion, Emission Control, Health Impacts, and Fuels Merit Review and Peer Evaluation, held May 15-18, 2006 at Argonne National Laboratory. The work evaluated in this document supports the FreedomCAR and Vehicle Technologies Program. The results of this merit review and peer evaluation are major inputs used by DOE in making its funding decisions for the upcoming fiscal year.

  11. Advanced chemical hydride-based hydrogen generation/storage system for fuel cell vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Breault, R.W.; Rolfe, J. [Thermo Power Corp., Waltham, MA (United States)

    1998-08-01

    Because of the inherent advantages of high efficiency, environmental acceptability, and high modularity, fuel cells are potentially attractive power supplies. Worldwide concerns over clean environments have revitalized research efforts on developing fuel cell vehicles (FCV). As a result of intensive research efforts, most of the subsystem technology for FCV`s are currently well established. These include: high power density PEM fuel cells, control systems, thermal management technology, and secondary power sources for hybrid operation. For mobile applications, however, supply of hydrogen or fuel for fuel cell operation poses a significant logistic problem. To supply high purity hydrogen for FCV operation, Thermo Power`s Advanced Technology Group is developing an advanced hydrogen storage technology. In this approach, a metal hydride/organic slurry is used as the hydrogen carrier and storage media. At the point of use, high purity hydrogen will be produced by reacting the metal hydride/organic slurry with water. In addition, Thermo Power has conceived the paths for recovery and regeneration of the spent hydride (practically metal hydroxide). The fluid-like nature of the spent hydride/organic slurry will provide a unique opportunity for pumping, transporting, and storing these materials. The final product of the program will be a user-friendly and relatively high energy storage density hydrogen supply system for fuel cell operation. In addition, the spent hydride can relatively easily be collected at the pumping station and regenerated utilizing renewable sources, such as biomass, natural, or coal, at the central processing plants. Therefore, the entire process will be economically favorable and environmentally friendly.

  12. Biofuels Fuels Technology Pathway Options for Advanced Drop-in Biofuels Production

    Energy Technology Data Exchange (ETDEWEB)

    Kevin L Kenney

    2011-09-01

    Advanced drop-in hydrocarbon biofuels require biofuel alternatives for refinery products other than gasoline. Candidate biofuels must have performance characteristics equivalent to conventional petroleum-based fuels. The technology pathways for biofuel alternatives also must be plausible, sustainable (e.g., positive energy balance, environmentally benign, etc.), and demonstrate a reasonable pathway to economic viability and end-user affordability. Viable biofuels technology pathways must address feedstock production and environmental issues through to the fuel or chemical end products. Potential end products include compatible replacement fuel products (e.g., gasoline, diesel, and JP8 and JP5 jet fuel) and other petroleum products or chemicals typically produced from a barrel of crude. Considering the complexity and technology diversity of a complete biofuels supply chain, no single entity or technology provider is capable of addressing in depth all aspects of any given pathway; however, all the necessary expert entities exist. As such, we propose the assembly of a team capable of conducting an in-depth technology pathway options analysis (including sustainability indicators and complete LCA) to identify and define the domestic biofuel pathways for a Green Fleet. This team is not only capable of conducting in-depth analyses on technology pathways, but collectively they are able to trouble shoot and/or engineer solutions that would give industrial technology providers the highest potential for success. Such a team would provide the greatest possible down-side protection for high-risk advanced drop-in biofuels procurement(s).

  13. Environmental assessment for decontaminating and decommissioning the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, PA

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The Department of Energy has prepared an environmental assessment on the proposed decontamination and decommissioning of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, Pennsylvania. Based on the environmental assessment, which is available to the public on request, the Department has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act of 1969, 42 USC 4321 et seq. Therefore, no environmental impact statement is required. The proposed action is to decontaminate and decommission the Westinghouse Advanced Reactors Division fuel fabrication facilities (the Plutonium Laboratory - Building 7, and the Advanced Fuels Laboratory - Building 8). Decontamination and decommissioning of the facilities would require removal of all process equipment, the associated service lines, and decontamination of the interior surfaces of the buildings so that the empty buildings could be released for unrestricted use. Radioactive waste generated during these activities would be transported in licensed containers by truck for disposal at the Department's facility at Hanford, Washington. Useable non-radioactive materials would be sold as excess material, and non-radioactive waste would be disposed of by burial as sanitary landfill at an approved site.

  14. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the 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 several independent 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  15. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  16. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    International Nuclear Information System (INIS)

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called 'AGR-1,' graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on fuel

  17. Detection of environmental radioactive microparticles bearing fissile material

    Energy Technology Data Exchange (ETDEWEB)

    Tamborini, G.; Dequincey, O.; Betti, M. [European Commission, Joint Research Centre, Institute of Transuranium Elements, Karlsruhe (Germany)

    2004-07-01

    Dispersion of actinides and/or fission products in the environment is mainly due to radioactive micro-particles. The determination of the isotopic composition of the major constituents and fissile material in environmental micro-particles is a fundamental fingerprint for the detection of their source term. In order to facilitate the localisation of interesting particles on samples with low density of actinide containing particle and their successive isotopic composition analysis, complementary techniques based on the fission track method and thermal ionisation mass spectrometry (FT-TIMS) are being developed. The main purpose is to use the solid state nuclear track detection (SSNTD) method for the localisation of interesting particles as well as a first estimation of their activity and content of fissile material. In the present work a procedure of rapid recognition of {sup 235}U-enriched particles using this technique is presented as a first step of an analytical procedure for environmental particle survey. (orig.)

  18. Recent advances in microbial production of fuels and chemicals using tools and strategies of systems metabolic engineering

    DEFF Research Database (Denmark)

    Cho, Changhee; Choi, So Young; Luo, Zi Wei;

    2015-01-01

    The advent of various systems metabolic engineering tools and strategies has enabled more sophisticated engineering of microorganisms for the production of industrially useful fuels and chemicals. Advances in systems metabolic engineering have been made in overproducing natural chemicals and prod...

  19. What should ''damaged'' mean in air transport of fissile packages

    International Nuclear Information System (INIS)

    It is likely that the ongoing process to produce the 1996 version of the IAEA Regulation for the Safe Transport of Radioactive Materials, IAEA Safety Series 6(SS 6) will result in a more stringent package qualification standard for air transport of large quantities of radioactive materials (RAM) than is included in the 1990 version. During the process to define the scope of the new requirements there was extensive discussion of their impact on, and application to, fissile material package qualification criteria. Since fissile materials are shipped in a variety of packagings ranging from exempt to Type B, each packaging of each type must be evaluated for its ability to maintain subcriticality both alone and in arrays and in both damaged and undamaged condition. In the 1990 version of SS 6 ''damaged'' means the condition of a package after it had undergone the ''tests for demonstrating the ability to withstand accident conditions in transport,'' i.e., Type B qualification tests. These tests conditions are typical of severe accidents in surface modes, but are less severe than air mode qualification test environments to be applied to Type C packages. As a result, questions arose about the need for a corresponding change in the 1996 SS 6 to define ''damaged'' to include the Type C test regime for criticality evaluations of fissile packages in air transport

  20. Advanced characterization of MIMAS MOX fuel microstructure to quantify the HBS formation

    Energy Technology Data Exchange (ETDEWEB)

    Bouloré, Antoine, E-mail: antoine.boulore@cea.fr [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Aufore, Laurence; Federici, Eric [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Blanpain, Patrick [AREVA NP SAS, 10 rue Juliette Récamier, F-69456 Lyon (France); Blachier, Rémi [EDF, SEPTEN, 12-14 Av. Dutrievoz, F-69628 Villeurbanne (France)

    2015-01-15

    Highlights: • An advanced characterization of MIMAS MOX fuel based only on fresh fuel pellet characterization. • A probabilistic approach to model the High Burnup Structure formation in oxide fuels. • Validation of the method by comparing to experimental data obtained on fuel irradiated in the Halden reactor. - Abstract: Fission gas behaviour in accidental situations is closely related to the location of fission gas before the accident. More precisely, most of the fission gas in intergranular position is released during the accident and HBS zones contribute a lot to this intergranular quantity. So a methodology to characterize the HBS zones a priori from examination of unirradiated pellet has been developed at CEA. Characterization of plutonium distribution in MIMAS MOX fresh fuel pellets can be performed by image analysis on 1 mm{sup 2} X-ray mappings of plutonium acquired using Electron Probe Micro Analysis (EPMA). The specific software developed to describe the fuel using Pu X-ray mapping (ANACONDA) has been improved in order to simulate the fission products (FP) production and recoil during a given irradiation of the fuel, taking into account the evolution of the plutonium due to neutron irradiation. This simulation results from calculations with our fuel performance code ALCYONE combined with image processing. The final result is a mapping of local burn-up, but also the distribution of the relative FP concentration as a function of the local burn-up. A validation of this simulation process has been done by comparing the simulated mapping of neodymium to one measured on the same fuel batch after irradiation. Using previous studies of mechanisms for HBS formation, a probabilistic criterion for HBS formation has been proposed, based on the EPMA measurements of the decrease of the xenon signal as a function of the local burn-up. Combining the simulated FP cartography with this probabilistic HBS formation criterion, it is possible to calculate the surface

  1. Novel approaches in advanced combustion characterization of fuels for advanced pressurized combustion

    Energy Technology Data Exchange (ETDEWEB)

    Aho, M.; Haemaelaeinen, J. [VTT Energy (Finland); Joutsenoja, T. [Tampere Univ. of Technology (Finland)

    1996-12-01

    This project is a part of the EU Joule 2 (extension) programme. The objective of the research of Technical Research Centre of Finland (VTT) is to produce experimental results of the effects of pressure and other important parameters on the combustion of pulverized coals and their char derivates. The results can be utilized in modelling of pressurized combustion and in planning pilot-scale reactors. The coals to be studied are Polish hvb coal, French lignite (Gardanne), German anthracite (Niederberg) and German (Goettelbom) hvb coal. The samples are combusted in an electrically heated, pressurized entrained flow reactor (PEFR), where the experimental conditions are controlled with a high precision. The particle size of the fuel can vary between 100 and 300 {mu}m. The studied things are combustion rates, temperatures and sizes of burning single coal and char particles. The latter measurements are performed with a method developed by Tampere University of Technology, Finland. In some of the experiments, mass loss and elemental composition of the char residue are studied in more details as the function of time to find out the combustion mechanism. Combustion rate of pulverized (140-180 {mu}m) Gardanne lignite and Niederberg anthracite were measured and compared with the data obtained earlier with Polish hvb coal at various pressures, gas temperatures, oxygen partial pressures and partial pressures of carbon dioxide in the second working period. In addition, particle temperatures were measured with anthracite. The experimental results were treated with multivariable partial least squares (PLS) method to find regression equation between the measured things and the experimental variables. (author)

  2. Advanced methods of process/quality control in nuclear reactor fuel manufacture. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Nuclear fuel plays an essential role in ensuring the competitiveness of nuclear energy and its acceptance by the public. The economic and market situation is not favorable at present for nuclear fuel designers and suppliers. The reduction in fuel prices (mainly to compete with fossil fuels) and in the number of fuel assemblies to be delivered to customers (mainly due to burnup increase) has been offset by the rising number of safety and other requirements, e.g. the choice of fuel and structural materials and the qualification of equipment. In this respect, higher burnup and thermal rates, longer fuel cycles and the use of MOX fuels are the real means to improve the economics of the nuclear fuel cycle as a whole. Therefore, utilities and fuel vendors have recently initiated new research and development programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel for safe and reliable reactor operation more demanding conditions. In this connection, improvement of fuel quality occupies an important place and this requires continuous effort on the part of fuel researchers, designers and producers. In the early years of commercial fuel fabrication, emphasis was given to advancements in quality control/quality assurance related mainly to the product itself. Now, the emphasis is transferred to improvements in process control and to implementation of overall total quality management (TQM) programmes. In the area of fuel quality control, statistical methods are now widely implemented, replacing 100% inspection. The IAEA, recognizing the importance of obtaining and maintaining high standards in fuel fabrication, has paid particular attention to this subject. In response to the rapid progress in development and implementation of advanced methods of process/quality control in nuclear fuel manufacture and on the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA conducted a

  3. Coated Particles Fuel Compact-General Purpose Heat Source for Advanced Radioisotope Power Systems

    Science.gov (United States)

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

    2003-01-01

    Coated Particles Fuel Compacts (CPFC) have recently been shown to offer performance advantage for use in Radioisotope Heater Units (RHUs) and design flexibility for integrating at high thermal efficiency with Stirling Engine converters, currently being considered for 100 We. Advanced Radioisotope Power Systems (ARPS). The particles in the compact consist of 238PuO2 fuel kernels with 5-μm thick PyC inner coating and a strong ZrC outer coating, whose thickness depends on the maximum fuel temperature during reentry, the fuel kernel diameter, and the fraction of helium gas released from the kernels and fully contained by the ZrC coating. In addition to containing the helium generated by radioactive decay of 238Pu for up to 10 years before launch and 10-15 years mission lifetime, the kernels are intentionally sized (>= 300 μm in diameter) to prevent any adverse radiological effects on reentry. This paper investigates the advantage of replacing the four iridium-clad 238PuO2 fuel pellets, the two floating graphite membranes, and the two graphite impact shells in current State-Of-The-Art (SOA) General Purpose Heat Source (GPHS) with CPFC. The total mass, thermal power, and specific power of the CPFC-GPHS are calculated as functions of the helium release fraction from the fuel kernels and maximum fuel temperature during reentry from 1500 K to 2400 K. For the same total mass and volume as SOA GPHS, the generated thermal power by single-size particles CPFC-GPHS is 260 W at Beginning-Of-Mission (BOM), versus 231 W for the GPHS. For an additional 10% increase in total mass, the CPFC-GPHS could generate 340 W BOM; 48% higher than SOA GPHS. The corresponding specific thermal power is 214 W/kg, versus 160 W/kg for SOA GPHS; a 34% increase. Therefore, for the same thermal power, the CPFC-GPHS is lighter than SOA GPHS, while it uses the same amount of 238PuO2 fuel and same aeroshell. For the same helium release fraction and fuel temperature, binary-size particles CPFC-GPHS could

  4. Recent advances in hardware and software are to improve spent fuel measurements

    International Nuclear Information System (INIS)

    Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's 3He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined for

  5. Safety related issues of spent nuclear fuel storage : summary of a NATO advanced research workshop

    International Nuclear Information System (INIS)

    Full text: A NATO Advanced Research Workshop was held in Almaty, Kazakhstan, in September 2005. The Workshop was co-sponsored by the IAEA and was concerned with the safety issues associated with spent fuel and waste from three types of reactor: research reactors with Al alloy-clad dispersion fuels, fast reactors with stainless steel-clad UO2, and commercial light-water reactors with Zr alloy-clad UO2. Fifteen presentations dealt with research reactors, five with the BN-350 fast reactor in Kazakhstan-shut down and in decommissioning, and two with commercial reactors in the U.S. and Ukraine. With 657 research reactors built and 274 still operational, corrosion of Al-clad research reactor spent fuel during wet storage was a major subject for discussion. Programs at the IAEA, in the U.S., and elsewhere, have actively studied corrosion of Al-clad fuel since the 1980s and the major mechanisms for aqueous corrosion of both spent fuel and of spent-fuel-pool structural components appear to be now well understood, as are the procedures required to limit corrosion. Nonetheless, avoiding corrosion requires vigilance in monitoring and controlling water quality. Measures to ensure water quality are now being taken at operating research reactors, but are difficult to impose at reactors that are shutdown, where there is less funding (or staff) for the task. It was noted there are about 62,000 spent research reactor fuel assemblies-most of them in wet storage-at many reactor sites around the world, three-quarters in industrialized nations, the remainder in developing countries. Dry storage of research reactor fuel is also being used or actively considered in France, Poland, Russia and the U.S. A variant of simple dry storage-the 'melt-and-dilute' option-casts the spent research reactor fuel with natural U into steel canisters to produce a corrosion-resistant low-enrichment fuel configuration which is suitable for safe long-term storage. The main safety issue of spent fast reactor

  6. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  7. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Satmark, B.; Apostolidis, C.; Courson, O.; Malmbeck, R.; Carlos, R.; Pagliosa, G.; Romer, K.; Glatz, J.P. [European Commission, DG-JRC, Institute for Transuranium Elements, Hot Cell Technology, Karlsruhe (Germany)

    2000-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products, lanthanides and the minor actinides (MA) were used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, Diamide and CYANEX 923 as extractant. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination, Diamide and BTP was found to be the best among extractants tested to achieve an efficient MA recovery from spent fuel. (authors)

  8. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, B.; Apostolidis, C.; Carlos, R.; Courson, O.; Glatz, J.P.; Malmbeck, R.; Pagliosa, G.; Roemer, K.; Serrano-Purroy, D. [European Commission, JRC, Inst. for Transuranium Elements, Karlsruhe (Germany)

    2004-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products including lanthanides and the minor actinides (MA) was used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, diamide and CYANEX 923 as extractants. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm separation. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination of diamide and BTP seems to be the best, among extractants tested, to achieve an efficient MA recovery from spent fuel. (orig.)

  9. The development of technical database of advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Ro, Seung Gy; Byeon, Kee Hoh; Song, Dae Yong; Park, Seong Won; Shin, Young Jun

    1999-03-01

    The purpose of this study is to develop the technical database system to provide useful information to researchers who study on the back end nuclear fuel cycle. Technical database of advanced spent fuel management process was developed for a prototype system in 1997. In 1998, this database system is improved into multi-user systems and appended special database which is composed of thermochemical formation data and reaction data. In this report, the detailed specification of our system design is described and the operating methods are illustrated as a user's manual. Also, expanding current system, or interfacing between this system and other system, this report is very useful as a reference. (Author). 10 refs., 18 tabs., 46 fig.

  10. Materials Research Advances towards High-Capacity Battery/Fuel Cell Devices (Invited paper)

    Institute of Scientific and Technical Information of China (English)

    Wei-Dong He; Lu-Han Ye; Ke-Chun Wen; Ya-Chun Liang; Wei-Qiang Lv; Gao-Long Zhu; Kelvin H. L. Zhang

    2016-01-01

    The world has entered an era featured with fast transportations, instant communications, and prompt technological revolutions, the further advancement of which all relies fundamentally, yet, on the development of cost-effective energy resources allowing for durable and high-rate energy supply. Current battery and fuel cell systems are challenged by a few issues characterized either by insufficient energy capacity or by operation instability and, thus, are not ideal for such highly-demanded applications as electrical vehicles and portable electronic devices. In this mini-review, we present, from materials perspectives, a few selected important breakthroughs in energy resources employed in these applications. Prospectives are then given to look towards future research activities for seeking viable materials solutions for addressing the capacity, durability, and cost shortcomings associated with current battery/fuel cell devices.

  11. The development of technical database of advanced spent fuel management process

    International Nuclear Information System (INIS)

    The purpose of this study is to develop the technical database system to provide useful information to researchers who study on the back end nuclear fuel cycle. Technical database of advanced spent fuel management process was developed for a prototype system in 1997. In 1998, this database system is improved into multi-user systems and appended special database which is composed of thermochemical formation data and reaction data. In this report, the detailed specification of our system design is described and the operating methods are illustrated as a user's manual. Also, expanding current system, or interfacing between this system and other system, this report is very useful as a reference. (Author). 10 refs., 18 tabs., 46 fig

  12. Advanced control approach for hybrid systems based on solid oxide fuel cells

    International Nuclear Information System (INIS)

    Highlights: • Advanced new control system for SOFC based hybrid plants. • Proportional–Integral approach with feed-forward technology. • Good control of fuel cell temperature. • All critical properties maintained inside safe conditions. - Abstract: This paper shows a new advanced control approach for operations in hybrid systems equipped with solid oxide fuel cell technology. This new tool, which combines feed-forward and standard proportional–integral techniques, controls the system during load changes avoiding failures and stress conditions detrimental to component life. This approach was selected to combine simplicity and good control performance. Moreover, the new approach presented in this paper eliminates the need for mass flow rate meters and other expensive probes, as usually required for a commercial plant. Compared to previous works, better performance is achieved in controlling fuel cell temperature (maximum gradient significantly lower than 3 K/min), reducing the pressure gap between cathode and anode sides (at least a 30% decrease during transient operations), and generating a higher safe margin (at least a 10% increase) for the Steam-to-Carbon Ratio. This new control system was developed and optimized using a hybrid system transient model implemented, validated and tested within previous works. The plant, comprising the coupling of a tubular solid oxide fuel cell stack with a microturbine, is equipped with a bypass valve able to connect the compressor outlet with the turbine inlet duct for rotational speed control. Following model development and tuning activities, several operative conditions were considered to show the new control system increased performance compared to previous tools (the same hybrid system model was used with the new control approach). Special attention was devoted to electrical load steps and ramps considering significant changes in ambient conditions

  13. Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

    International Nuclear Information System (INIS)

    scenarios and breakdown of LCOE contributors rather than provision of absolute values, as technological readiness levels are low for most of the advanced fuel cycle stages. The obtained estimations show an increase of LCOE – averaged over the whole period – with respect to the reference open cycle scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios. The main contribution to LCOE is the capital costs of new facilities, quantified between 60% and 69% depending on the scenario. An uncertainty analysis is provided around assumed low and high values of processes and technologies

  14. Establishment of team work system for advanced spent fuel management process

    International Nuclear Information System (INIS)

    The advanced spent fuel management process (ASFMP), which is being developed by KAERI, is now in the 2nd research phase. This phase has a goal to design the total system of active demonstration of ASFMP. It is composed of the core process, remote handling technologies, examination technologies and experimental facilities. For the collaboration of these research fields, a team work system has been established by proper hardware and software selections for use of about 50 project members. This system has been tested by adaptation to the ASFMP project and will be used during the remained project period

  15. Development of an advanced bond coat for solid oxide fuel cell interconnector applications

    Science.gov (United States)

    Yeh, An-Chou; Chen, Yu-Ming; Liu, Chien-Kuo; Shong, Wei-Ja

    2015-11-01

    An advanced bond coat has been developed for solid oxide fuel cell interconnector applications; a low thermal expansion superalloy has been selected as the substrate, and the newly developed bond coat is applied between the substrate and the LSM top coat. The bond coat composition is designed to be near thermodynamic equilibrium with the substrate to minimize interdiffusion with the substrate while providing oxidation protection for the substrate. The bond coat exhibits good oxidation resistance, a low area specific resistance, and a low thermal expansion coefficient at 800 °C; experimental results indicate that interdiffusion between the bond coat and the substrate can be hindered.

  16. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D-3He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D-3He tokamak plasma

  17. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  18. Monolithic solid oxide fuel cell technology advancement for coal-based power generation

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-14

    The program is conducted by a team consisting of AiResearch Los Angeles Division of Allied-Signal Aerospace Company and Argonne National Laboratory (ANL). The objective of the program is to advance materials and fabrication methodologies to develop a monolithic solid oxide fuel cell (MSOFC) system capable of meeting performance, life, and cost goals for coal-based power generation. The program focuses on materials research and development, fabrication process development, cell/stack performance testing and characterization, cost and system analysis, and quality development.

  19. BOR-60 reactor as an instrument for experimental substantiation of fuel rods for advanced NPPs

    International Nuclear Information System (INIS)

    Full text: The BOR-60 fast test reactor is actually the only facility of this type in the world that has been in reliable and continuous operation for about 35 years. One of the principle reactor tasks is irradiation of advanced fuel and structural materials in different conditions. Inside the reactor the materials can be irradiated in any core and reflector cell except seven cells used for control rods. The number of fuel assemblies loaded into the reactor can vary from 85 to 124 depending on the burnup, core configuration and fuel properties. Due to the reactor design, the core dimensions can be widely changed allowing accommodation of no less than 20 experimental assemblies in different reactor cells. The neutron flux value in individual cells can vary more than 3 times at the maximum value of 3.7·1015 n/cm2s. Thus various fuel compositions can be loaded into the reactor and practically any burnups can achieve. Based on the long-term investigation of the reactor characteristics, we studied the reactor behavior in different conditions, developed a set of the verified codes and different procedures for the on-line reactor maintenance and performance of the wide scope of experiments. A set of specialized testing facilities consisting of capsule units and dismountable assemblies are used for irradiation of the wide range of materials and items at different conditions. The advantages of these facilities are their simplicity and possibility of installation in any core and reflector cell. In addition to the precision calculations of the irradiation conditions there is also a possibility for monitoring the neutron flux and temperature. A special thermometric channel available in the core allows accommodation of the experimental facilities and output of information of the irradiation conditions by 30-50 communication lines. It was required to develop a series of independent instrumented capsule-loops, special instrumented fuel assemblies etc. to be used in the channel

  20. Development of safeguards technology for lab-scale advanced fuel cycle facility at KAERI

    International Nuclear Information System (INIS)

    KAERI (Korea Atomic Energy Research Institute) has been developing the DUPIC (Direct Use of PWR spent fuel in CANDU) fuel cycle and ACP (Advanced Spent Fuel Conditioning Process) technology for the purpose of spent fuel management. A safeguards system has been applied to R and D process for fabricating DUPIC fuel directly with PWR spent fuel material. Safeguards issues to be resolved were identified in the areas such as international cooperation on handling foreign origin nuclear material, technology development of operator's measurement system of bulk handling process of spent fuel material, and built-in C/S system for independent verification of material flow. All those safeguards issues have been finally resolved. The lab-scale DUPIC facility (DFDF) safeguards system was successfully established under the international cooperation program. The ACP has been under development at KAERI since 1997 to tackle the problem of the accumulation of the spent fuel. The concept is to convert the spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat power, volume, and radioactivity of the spent fuel. The main objective of the ACP is to treat the PWR spent fuel for a long-term storage and eventual disposal in a proliferation resistant and cost effective way. Moreover, the electrolytic reduction method of the ACP can contribute to the innovative nuclear energy system as a key technology for the preparation of the metallic fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in the ACP facility (ACPF) to validate the concept. Based on the results of a safeguards implementation at DFDF hot cell, the reference safeguards design conditions are established for the ACPF. Basically, the nuclear material accounting will be performed by ASNC (ACP Safeguards Neutron Counter), which is the same concept as the

  1. Diesel engine emissions and combustion predictions using advanced mixing models applicable to fuel sprays

    Science.gov (United States)

    Abani, Neerav; Reitz, Rolf D.

    2010-09-01

    An advanced mixing model was applied to study engine emissions and combustion with different injection strategies ranging from multiple injections, early injection and grouped-hole nozzle injection in light and heavy duty diesel engines. The model was implemented in the KIVA-CHEMKIN engine combustion code and simulations were conducted at different mesh resolutions. The model was compared with the standard KIVA spray model that uses the Lagrangian-Drop and Eulerian-Fluid (LDEF) approach, and a Gas Jet spray model that improves predictions of liquid sprays. A Vapor Particle Method (VPM) is introduced that accounts for sub-grid scale mixing of fuel vapor and more accurately and predicts the mixing of fuel-vapor over a range of mesh resolutions. The fuel vapor is transported as particles until a certain distance from nozzle is reached where the local jet half-width is adequately resolved by the local mesh scale. Within this distance the vapor particle is transported while releasing fuel vapor locally, as determined by a weighting factor. The VPM model more accurately predicts fuel-vapor penetrations for early cycle injections and flame lift-off lengths for late cycle injections. Engine combustion computations show that as compared to the standard KIVA and Gas Jet spray models, the VPM spray model improves predictions of in-cylinder pressure, heat released rate and engine emissions of NOx, CO and soot with coarse mesh resolutions. The VPM spray model is thus a good tool for efficiently investigating diesel engine combustion with practical mesh resolutions, thereby saving computer time.

  2. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

  3. Feasibility Study on AFR-100 Fuel Conversion from Uranium-based Fuel to Thorium-based Fuel

    International Nuclear Information System (INIS)

    The feasibility study of converting a fast reactor from uranium-based fuel to thorium-based fuel was studied using the 100 MWe Advanced Fast Reactor (AFR-100). Several fuel conversion scenarios were envisioned in this study. The first scenario is a progressive fuel conversion without fissile support. It consists of progressively replacing the burnt uranium-based fuel with pure thorium-based fuel without fissile material addition. This was found to be impractical because the low excess reactivity of the uranium-fuelled AFR-100 core, resulting in an extremely short cycle length even when only a few assemblies are replaced. A second scenario consists in operating the reference LEU fuelled AFR-100 core for 24 years and then replacing one fuel batch out of four every 7.04 years with thorium-based fuel mixed with transuranics. The transuranics weight fraction required during the transition period is identical to that required at equilibrium and is equal to 18.6%. The original uranium-based fuel is discharged with an average burnup of 120 GWd/t and the Th-TRU fuel with an average burnup of 101 GWd/t. The thermal-hydraulic and passive safety performances of this core are similar to those of the reference AFR-100 design. However, Th-TRU fuel fabrication and performance needs to be demonstrated and TRU separation from the LWR used nuclear fuel is necessary. The third scenario proposed consists of replacing the whole AFR-100 core with fuel assemblies made of several thorium and 20% enriched LEU layers. The mode of operation is similar to that of the reference AFR-100 core with the exception of the cycle length which needs to be reduced from 30 to 18 years. The average LEU and thorium discharge burnups are 79 GWd/t and 23 GWd/t, respectively. The major benefit of this approach is the improved inherent safety of the reactor due to the reduced coolant void worth. (author)

  4. Integrated safeguards testing laboratories in support of the advanced fuel cycle initiative

    Energy Technology Data Exchange (ETDEWEB)

    Santi, Peter A [Los Alamos National Laboratory; Demuth, Scott F [Los Alamos National Laboratory; Klasky, Kristen L [Los Alamos National Laboratory; Lee, Haeok [Los Alamos National Laboratory; Miller, Michael C [Los Alamos National Laboratory; Sprinkle, James K [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Williams, Bradley [DOE, NE

    2009-01-01

    A key enabler for advanced fuel cycle safeguards research and technology development for programs such as the Advanced Fuel Cycle Initiative (AFCI) is access to facilities and nuclear materials. This access is necessary in many cases in order to ensure that advanced safeguards techniques and technologies meet the measurement needs for which they were designed. One such crucial facility is a hot cell based laboratory which would allow developers from universities, national laboratories, and commercial companies to perform iterative research and development of advanced safeguards instrumentation under realistic operating conditions but not be subject to production schedule limitations. The need for such a facility arises from the requirement to accurately measure minor actinide and/or fission product bearing nuclear materials that cannot be adequately shielded in glove boxes. With the contraction of the DOE nuclear complex following the end of the cold war, many suitable facilities at DOE sites are increasingly costly to operate and are being evaluated for closure. A hot cell based laboratory that allowed developers to install and remove instrumentation from the hot cell would allow for both risk mitigation and performance optimization of the instrumentation prior to fielding equipment in facilities where maintenance and repair of the instrumentation is difficult or impossible. These benefits are accomplished by providing developers the opportunity to iterate between testing the performance of the instrumentation by measuring realistic types and amounts of nuclear material, and adjusting and refining the instrumentation based on the results of these measurements. In this paper, we review the requirements for such a facility using the Wing 9 hot cells in the Los Alamos National Laboratory's Chemistry and Metallurgy Research facility as a model for such a facility and describe recent use of these hot cells in support of AFCI.

  5. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6Li in order to reduce parasite neutron captures in there. (orig./HP)

  6. Fissile material disposition program final immobilization form assessment and recommendation

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL), in its role as the lead laboratory for the development of plutonium immobilization technologies for the Department of Energy's Office of Fissile Materials Disposition (MD), has been requested by MD to recommend an immobilization technology for the disposition of surplus weapons- usable plutonium. The recommendation and supporting documentation was requested to be provided by September 1, 1997. This report addresses the choice between glass and ceramic technologies for immobilizing plutonium using the can-in-canister approach. Its purpose is to provide a comparative evaluation of the two candidate technologies and to recommend a form based on technical considerations

  7. Engineering development of advanced physical fine coal cleaning for premium fuel applications

    International Nuclear Information System (INIS)

    Bechtel, together with Amax Research and Development Center (Amax R ampersand D), has prepared this study which provides conceptual cost estimates for the production of premium quality coal-water slurry fuel (CWF) in a commercial plant. Two scenarios are presented, one using column flotation technology and the other the selective agglomeration to clean the coal to the required quality specifications. This study forms part of US Department of Energy program Engineering Development of Advanced Physical Fine Coal Cleaning for Premium Fuel Applications, (Contract No. DE-AC22- 92PC92208), under Task 11, Project Final Report. The primary objective of the Department of Energy program is to develop the design base for prototype commercial advanced fine coal cleaning facilities capable of producing ultra-clean coals suitable for conversion to stable and highly loaded CWF. The fuels should contain less than 2 lb ash/MBtu (860 grams ash/GJ) of HHV and preferably less than 1 lb ash/MBtu (430 grams ash/GJ). The advanced fine coal cleaning technologies to be employed are advanced column froth flotation and selective agglomeration. It is further stipulated that operating conditions during the advanced cleaning process should recover not less than 80 percent of the carbon content (heating value) in the run-of-mine source coal. These goals for ultra-clean coal quality are to be met under the constraint that annualized coal production costs does not exceed $2.5 /MBtu ($ 2.37/GJ), including the mine mouth cost of the raw coal. A further objective of the program is to determine the distribution of a selected suite of eleven toxic trace elements between product CWF and the refuse stream of the cleaning processes. Laboratory, bench-scale and Process Development Unit (PDU) tests to evaluate advanced column flotation and selective agglomeration were completed earlier under this program with selected coal samples. A PDU with a capacity of 2 st/h was designed by Bechtel and installed at

  8. Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method

    Science.gov (United States)

    Metzger, Kathryn E.

    In light of the disaster at the Fukushima Daiichi Nuclear Plant, the Department of Energy's Advanced Fuels Program has shifted its interest from enhanced performance fuels to enhanced accident tolerance fuels. Dispersion fuels possess higher thermal conductivities than traditional light water reactor fuel and as a result, offer improved safety margins. The benefits of a dispersion fuel are due to the presence of the secondary non-fissile phase (matrix), which serves as a barrier to fission products and improves the overall thermal performance of the fuel. However, the presence of a matrix material reduces the fuel volume, which lowers the fissile content of dispersion. This issue can be remedied through the development of higher density fuel phases or through an optimization of fuel particle size and volume loading. The latter requirement necessitates the development of fabrication methods to produce small, micron-order fuel particles. This research examines the capabilities of the spark erosion process to fabricate particles on the order of 10 μm. A custom-built spark erosion device by CT Electromechanica was used to produce stainless steel surrogate fuel particles in a deionized water dielectric. Three arc intensities were evaluated to determine the effect on particle size. Particles were filtered from the dielectric using a polycarbonate membrane filter and vacuum filtration system. Fabricated particles were characterized via field emission scanning electron microscopy (FESEM), laser light particle size analysis, energy-dispersive spectroscopy (EDS), X-ray diffraction analysis (XRD), and gas pycnometry. FESEM images reveal that the spark erosion process produces highly spherical particles on the order of 10 microns. These findings are substantiated by the results of particle size analysis. Additionally, EDS and XRD results indicate the presence of oxide phases, which suggests the dielectric reacted with the molten debris during particle formation.

  9. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  10. Engineering Development of Advanced Physical Fine Coal Cleaning for Premium Fuel Applications

    Energy Technology Data Exchange (ETDEWEB)

    Smit, Frank J; Schields, Gene L; Jha, Mehesh C; Moro, Nick

    1997-09-26

    The ash in six common bituminous coals, Taggart, Winifrede, Elkhorn No. 3, Indiana VII, Sunnyside and Hiawatha, could be liberated by fine grinding to allow preparation of clean coal meeting premium fuel specifications (< 1- 2 lb/ MBtu ash and <0.6 lb/ MBtu sulfur) by laboratory and bench- scale column flotation or selective agglomeration. Over 2,100 tons of coal were cleaned in the PDU at feed rates between 2,500 and 6,000 lb/ h by Microcel™ column flotation and by selective agglomeration using recycled heptane as the bridging liquid. Parametric testing of each process and 72- hr productions runs were completed on each of the three test coals. The following results were achieved after optimization of the operating parameters: The primary objective was to develop the design base for commercial fine coal cleaning facilities for producing ultra- clean coals which can be converted into coal-water slurry premium fuel. The coal cleaning technologies to be developed were advanced column flotation and selective agglomeration, and the goal was to produce fuel meeting the following specifications.

  11. Use of AWCC in evaluation of unknown fissile materials

    International Nuclear Information System (INIS)

    An important technological problem in the sphere of non-proliferation and safeguards is nondestructive analysis (NDA) methodology for qualitative and quantitative characterization of nuclear materials. Additionally, NDA tends to be labor and time intensive. Two NDA techniques used at KIPT included Gamma Spectroscopy (for qualitative analysis) and Neutron activation (for quantitative analysis). Gamma Spectroscopy was used to confirm the presence of radionuclides within the samples, whereas an Active Well Coincidence Counter (AWCC) in the active mode was used to quantitatively determine the 235U content in particular types of fissile materials at KIPT. This paper describes the usage of the AWCC at NSG KIPT for characterizing nuclear materials for IAEA safeguards. It was also an opportunity to estimate fissile materials of unknown composition. The equipment used was a model JCC-51 AWCC using a shift register model JSR-12 from Canberra and two neutron sources [AN-HP (241ArnLi)]. A Compaq Presario computer using Windows version of NCC (Los Alamos software) was used to operate the AWCC. Materials studied in this project included highly enriched nuclear material in the form of powder, compacts (tablets, microspheres, rods), salt and scrap. The chemical composition of nuclear material included uranium metal, uranium dioxide, uranium nitride, uranium carbonitride, thorium dioxide, and mixtures of these compounds. Scrap consisted of uranium and impurities of hydrocarbons, carbon, silicon, tungsten, etc

  12. Advanced Simulation of Fuel Behavior Under Irradiation in the Pleiades Software Environment

    International Nuclear Information System (INIS)

    A “multi design” new generation software environment called PLEIADES has been developed by the CEA in the framework of a research cooperative program with EDF and AREVA. In this general software environment, ALCYONE is the PWR fuel performance simulation code. It is a multi-dimensional simulation software (1D, 2D and 3D), with applications for normal, transient and accidental conditions. It also has several levels of modelling, from industrial models to mechanistic ones depending on the amount of multi-scale details expected in the results of the simulation. The different dimensional schemes share the same thermomechanical Finite Element Method code CAST3M. The 1D scheme describes the behaviour of the whole rod and gives access to integral values such as rod fission gas release, clad profilometry and elongation. The 3D scheme allows a local study of Pellet Clad Mechanical Interaction (PCMI) by modelling the thermo-mechanical behaviour of one or several pellet fragments and overlying cladding. The 2D scheme is a compromise between calculation time and the accuracy of the local fuel description. Recently the 3D approach has been extended to a short fuel rod model in order to simulate the ballooning phenomenon during accidental transients. In this paper, we will present the general description of the ALCYONE simulation code in the PLEIADES environment (general computation algorithm, advanced fission gas model for UO2 and MOX fuels, 3D computation scheme). A focus will be presented on specific developments which have already been done to simulate accidental conditions such as LOCA and fast transients for different dimensional models. (author)

  13. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  14. Fissile materials from nuclear arms reductions: A question of disposition

    International Nuclear Information System (INIS)

    This Session, 35T-2, of the Annual Meeting of the American Association for the Advancement of Science (AAAS) was held on February 18, 1991. The papers presented during this session covered a variety of issues and technologies concerning the disposition of the highly enriched uranium and plutonium salvaged from retired nuclear warheads. However, circumstances, including the amount of time available for the session, imposed limitations on the number and breadth of these papers. A comprehensive study of this topic should include a broader range of papers. This session included a paper on molten salt reactors designed to use highly enriched uranium or plutonium as fuel. Other options for the disposal of plutonium, such as transmutation using accelerators and underground vitrification using nuclear explosions, were not discussed during this session, but need to be considered. Individual papers are indexed separately

  15. Fusion hybrids for generation of advanced (231Pa+232U+233U+234U)-fuel in closed (U-Pu-Th)-fuel cycle

    International Nuclear Information System (INIS)

    Technology of controlled thermonuclear fusion (CTF) is traditionally regarded as a practically inexhaustible energy source. However, development, mastering, broad deployment of fast breeder reactors and closure of nuclear fuel cycle (NFC) can also extend fuel base of nuclear power industry (NPI) up to practically unlimited scales. Under these conditions, it seems reasonable to introduce into a circle of the CTF-related studies the works directed towards solving some principal problems which can appear in a large-scale NPI in closed NFC. The first challenge is a large scale of operations in NFC back-end that should be reduced by achieving substantially higher fuel burn-up in power nuclear reactors. The use of 231Pa-232Th-232U-233U fuel in light-water reactor (LWR) opens a possibility of principle to reach very high (about 30% HM) or even ultra-high fuel burn-up. The second challenge is a potential unauthorized proliferation of fissionable materials. As is known, a certain remarkable quantity of 232U being introduced into uranium fraction of nuclear fuel can produce a serious barrier against switching the fuel over to non-energy purposes. Involvement of hybrid thermonuclear reactors (HTR) into NPI structure can substantially facilitate resolving these problems. If HTR will be involved into NPI structure, then main HTR mission consists not in energy generation but in production of nuclear fuel with a certain isotope composition. The present paper analyzes some neutron-physical features in production of advanced nuclear fuels in thorium HTR blankets. The obtained results demonstrated that such a nuclear fuel may be characterized by very stable neutron-multiplying properties during full LWR operation cycle and by enhanced proliferation resistance too. The paper evaluates potential benefits from involvement of HTR with thorium blanket into the international closed NFC. (author)

  16. Advanced Fuel Cycle Initiative AFC-1D, AFC-1G, and AFC-1H End of FY-07 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Debra J Utterbeck; Gray S Chang; Misit A Lillo

    2007-09-01

    The purpose of the U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), is to develop and demonstrate the technologies needed to transmute the long-lived transuranic isotopes contained in spent nuclear fuel into shorter-lived fission products. Success in this undertaking could potentially dramatically decrease the volume of material requiring disposal with attendant reductions in long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is investigation of irradiation/transmutation effects on actinide-bearing metallic fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Goals of this initiative include addressing the limited irradiation performance data available on metallic fuels with high concentrations of Pu, Np and Am, as are envisioned for use as actinide transmutation fuels. The AFC-1 irradiation experiments of transmutation fuels are expected to provide irradiation performance data on non-fertile and low-fertile fuel forms specifically, irradiation growth and swelling, helium production, fission gas release, fission product and fuel constituent migration, fuel phase equilibria, and fuel-cladding chemical interaction. Contained in this report are the to-date physics evaluations performed on three of the AFC-1 experiments; AFC-1D, AFC-1G and AFC-1H. The AFC-1D irradiation experiment consists of metallic non-fertile fuel compositions with minor actinides for potential use in accelerator driven systems and AFC-1G and AFC-1H irradiation experiments are part of the fast neutron reactor fuel development effort. The metallic fuel experiments and nitride experiment are high burnup analogs to previously irradiated experiments and are to be irradiated to = 40 at.% burnup.

  17. Application of phase equilibria and chemical thermodynamics to the preparation, farbiration, and performance of advanced fast reactor fuel materials

    International Nuclear Information System (INIS)

    Described are some phase equilibria and chemical thermodynamics of systems relevant to the production and operation of the so-called ''advanced'' fast breeder reactor fuels. The systems discussed include UPu carbides, nitrides, oxycarbides and carbonitrides. Some examples of the application of these phase equilibria to the preparation, fabrication and behaviour of the materials in temperature gradients appropriate to reactor conditions are presented. Finally, aspects of the complex four and five component, U-C-O-N and U-Pu-C-O-N systems are discussed, a detailed knowledge of which is required for an analysis of advanced fuel behaviour

  18. PICFEE - Programme of integration of elementary fission curves while taking the evolution of fissile nuclides into account

    International Nuclear Information System (INIS)

    The author presents and discusses the use of the PICFEE code (presentation of the addressed physical problem, mathematical formulation, programming aspects) which, in the case of a nuclear fuel made of fissile elements such as 235U, 238U and 239Pu and having been submitted to any irradiation, allows the calculation of the variation in time of any quantity which was instantly proportional to the number of fissions by time unit (i.e. residual power, so on) after irradiation. Irradiation is defined by a succession of linear diagrams, separated or not by null power thresholds. The curve of the considered value is integrated on these thresholds. The code can also take the evolution of the fuel during its irradiation into account

  19. New methods development for SCORPIO-VVER core monitoring systems to address advanced VVER 440 fuel types

    International Nuclear Information System (INIS)

    With introduction of advanced design fuel with Gd burnable absorber to Czech and Slovak VVER 440 reactors SCORPIO-VVER CMS faces new requirements and challenges. New methodology and tools had to be developed in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics) to properly model and address new design features of Gadolinium bearing fuel of Gd 1 and Gd 2 type. These methods have to be adapted for implementation in SCORPIO-VVER CMS. The paper provides comprehensive list of requirements and open questions, which need to be properly addressed and clearly defined prior to major innovation of the system commence. All related fields are being step by step re-evaluated (neutron physics, thermal-hydraulics and fuel thermal-mechanics). Pin power determination methodology had to be improved. Higher geometrical complexity on upper and lower core ends, particularly for transition cores with different fuel types, led to change in axial nodalization. More stringent fuel related limits (design criteria, lover margins) with higher burn-up and required unit (fuel) maneuverability require new calculation strategy in fuel conditioning/de-conditioning PCI supervising module PES. Validation of simplified simulator for HiBu domain is under development using FEMAXI code and new 2D-3D tool development for CEZ utility. Conclusions of the paper concentrate on mid term and long term innovation plans for core and fuel operation reliability crucial systems. (Author)

  20. Determination of Basic Structure-Property Relations for Processing and Modeling in Advanced Nuclear Fuel: Microstructure Evolution and Mechanical Properties

    International Nuclear Information System (INIS)

    The project objective is to study structure-property relations in solid solutions of nitrides and oxides with surrogate elements to simulate the behavior of fuels of inert matrix fuels of interest to the Advanced Fuel Cycle Initiative (AFCI), with emphasis in zirconium-based materials. Work with actual fuels will be carried out in parallel in collaboration with Los Alamos National Laboratory (LANL). Three key aspects will be explored: microstructure characterization through measurement of global texture evolution and local crystallographic variations using Electron Backscattering Diffraction (EBSD); determination of mechanical properties, including fracture toughness, quasi-static compression strength, and hardness, as functions of load and temperature, and, finally, development of structure-property relations to describe mechanical behavior of the fuels based on experimental data. Materials tested will be characterized to identify the mechanisms of deformation and fracture and their relationship to microstructure and its evolution. New aspects of this research are the inclusion of crystallographic information into the evaluation of fuel performance and the incorporation of statistical variations of microstructural variables into simplified models of mechanical behavior of fuels that account explicitly for these variations. The work is expected to provide insight into processing conditions leading to better fuel performance and structural reliability during manufacturing and service, as well as providing a simplified testing model for future fuel production

  1. Use of AWCC in evaluation of unknown fissile materials

    International Nuclear Information System (INIS)

    An important technological problem in the sphere of non-proliferation and safeguards is nondestructive analysis (NDA) methodology for qualitative and quantitative characterization of nuclear materials. Additionally, NDA tends to be labor and time intensive. Two NDA techniques used at KIPT included Gamma Spectroscopy (for qualitative analysis) and Neutron activation (for quantitative analysis). Gamma Spectroscopy was used to confirm the presence of radionuclides within the samples, whereas an Active Well Coincidence Counter (AWCC) in the active mode was used to quantitatively determine the 235U content in particular types of fissile materials at KIPT. This paper describes the usage of the AWCC at NSC KIPT for characterizing nuclear materials for IAEA safeguards. It was also an opportunity to estimate fissile materials of unknown composition. The equipment used was a model JCC-51 AWCC using a shift register model JSR-12 from Canberra and two neutron sources [AN-HP (241AmLi)]. A Compaq Presario computer using Windows version of NCC (Los Alamos software) was used to operate the AWCC. Materials studied in this project included highly enriched nuclear material in the form of powder, compacts (tablets, microspheres, rods), salt and scrap. The chemical composition of nuclear material included uranium metal, uranium dioxide, uranium nitride, uranium carbonitride, thorium dioxide, and mixtures of these compounds. New standards were used to recalibrate the AWCC using material obtained at KIPT with documented compositional values more similar to the materials to be measured, and in similar containers as well. During this study, the AWCC calibration curves were obtained for uranium metal and uranium dioxide with different enrichments up to 90 % for 235U. A broad spectrum of other fissile materials of unknown composition with differing enrichments has also been studied, and the items' isotopic and quantitative compositions have also been determined. Using the new calibration

  2. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  3. PEM Fuel Cells with Bio-Ethanol Processor Systems A Multidisciplinary Study of Modelling, Simulation, Fault Diagnosis and Advanced Control

    CERN Document Server

    Feroldi, Diego; Outbib, Rachid

    2012-01-01

    An apparently appropriate control scheme for PEM fuel cells may actually lead to an inoperable plant when it is connected to other unit operations in a process with recycle streams and energy integration. PEM Fuel Cells with Bio-Ethanol Processor Systems presents a control system design that provides basic regulation of the hydrogen production process with PEM fuel cells. It then goes on to construct a fault diagnosis system to improve plant safety above this control structure. PEM Fuel Cells with Bio-Ethanol Processor Systems is divided into two parts: the first covers fuel cells and the second discusses plants for hydrogen production from bio-ethanol to feed PEM fuel cells. Both parts give detailed analyses of modeling, simulation, advanced control, and fault diagnosis. They give an extensive, in-depth discussion of the problems that can occur in fuel cell systems and propose a way to control these systems through advanced control algorithms. A significant part of the book is also given over to computer-aid...

  4. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

    International Nuclear Information System (INIS)

    A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

  5. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories, P.O. Box 5800 MS 0736, Albuquerque, NM 87185 (United States); Blink, James [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Carter, Joe [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States); Fratoni, Massimiliano; Greenberg, Harris; Sutton, Mark [Lawrence Livermore National Laboratory (United States); Howard, Robert [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

    2012-07-01

    A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

  6. Measurement of uranium and plutonium content in a fuel assembly using the RPI spent fuel assay device

    International Nuclear Information System (INIS)

    In this paper we report measurements of the significant parameters, the sensitivities of the slowing-down-time assay device to the fissile contents of a boiling water reactor (BWR) assembly mock-up of fresh fuel

  7. Nanostructured Ion-Exchange Membranes for Fuel Cells: Recent Advances and Perspectives.

    Science.gov (United States)

    He, Guangwei; Li, Zhen; Zhao, Jing; Wang, Shaofei; Wu, Hong; Guiver, Michael D; Jiang, Zhongyi

    2015-09-23

    Polymer-based materials with tunable nanoscale structures and associated microenvironments hold great promise as next-generation ion-exchange membranes (IEMs) for acid or alkaline fuel cells. Understanding the relationships between nanostructure, physical and chemical microenvironment, and ion-transport properties are critical to the rational design and development of IEMs. These matters are addressed here by discussing representative and important advances since 2011, with particular emphasis on aromatic-polymer-based nanostructured IEMs, which are broadly divided into nanostructured polymer membranes and nanostructured polymer-filler composite membranes. For each category of membrane, the core factors that influence the physical and chemical microenvironments of the ion nanochannels are summarized. In addition, a brief perspective on the possible future directions of nanostructured IEMs is presented. PMID:26270555

  8. Advances and recent trends in heterogeneous photo(electro)-catalysis for solar fuels and chemicals.

    Science.gov (United States)

    Highfield, James

    2015-01-01

    In the context of a future renewable energy system based on hydrogen storage as energy-dense liquid alcohols co-synthesized from recycled CO2, this article reviews advances in photocatalysis and photoelectrocatalysis that exploit solar (photonic) primary energy in relevant endergonic processes, viz., H2 generation by water splitting, bio-oxygenate photoreforming, and artificial photosynthesis (CO2 reduction). Attainment of the efficiency (>10%) mandated for viable techno-economics (USD 2.00-4.00 per kg H2) and implementation on a global scale hinges on the development of photo(electro)catalysts and co-catalysts composed of earth-abundant elements offering visible-light-driven charge separation and surface redox chemistry in high quantum yield, while retaining the chemical and photo-stability typical of titanium dioxide, a ubiquitous oxide semiconductor and performance "benchmark". The dye-sensitized TiO2 solar cell and multi-junction Si are key "voltage-biasing" components in hybrid photovoltaic/photoelectrochemical (PV/PEC) devices that currently lead the field in performance. Prospects and limitations of visible-absorbing particulates, e.g., nanotextured crystalline α-Fe2O3, g-C3N4, and TiO2 sensitized by C/N-based dopants, multilayer composites, and plasmonic metals, are also considered. An interesting trend in water splitting is towards hydrogen peroxide as a solar fuel and value-added green reagent. Fundamental and technical hurdles impeding the advance towards pre-commercial solar fuels demonstration units are considered. PMID:25884553

  9. The scaling of economic and performance parameters of DT and advanced fuel fusion reactors

    International Nuclear Information System (INIS)

    In this study, the plasma stability index beta and the fusion power density in the plasma were treated as independent variables to determine how they influenced three economic performance parameters of fusion reactors burning the DT and four advanced fusion fuel cycles. The economic/performance parameters included the total power produced per unit length of reactor; the mass per unit length, and the specific mass in kilograms/kilowatt. The scaling of these parameters with beta and fusion power density was examined for a common set of engineering assumptions on the allowable wall loading limits, the maximum magnetic field existing in the plasma, average blanket mass density, etc. It was found that the power per unit length decreased as the plasma power density and beta increased. This is a consequence of the fact that the first wall is a bottleneck in the energy flow from the plasma to the generating equipment, and the wall power flux will exceed wall loading limits if the plasma radius exceeds a critical value. If one wished to build an engineering test reactor which produced a burning plasma at the lowest possible initial cost, and without regard to whether such a reactor would ultimately produce the cheapest power, then one would minimize the mass per unit length. The mass per unit length decreases with increasing plasma power density and beta, with the DT reaction being the most expensive at a fixed plasma power density (because of its thicker blanket), and the least expensive at a fixed value of beta, at least up to values of beta of 50%. The specific mass, in kg/kw, which is a rough measure of the cost of the power generated by the reactor, shows an opposite trend. It increases with increasing plasma power density and beta. At a given plasma power density and low beta, the DT reaction gives the lowest specific mass, but at a fixed beta above 10%, the advanced fuel cycles have the lowest specific mass

  10. Advances and Recent Trends in Heterogeneous Photo(Electro-Catalysis for Solar Fuels and Chemicals

    Directory of Open Access Journals (Sweden)

    James Highfield

    2015-04-01

    Full Text Available In the context of a future renewable energy system based on hydrogen storage as energy-dense liquid alcohols co-synthesized from recycled CO2, this article reviews advances in photocatalysis and photoelectrocatalysis that exploit solar (photonic primary energy in relevant endergonic processes, viz., H2 generation by water splitting, bio-oxygenate photoreforming, and artificial photosynthesis (CO2 reduction. Attainment of the efficiency (>10% mandated for viable techno-economics (USD 2.00–4.00 per kg H2 and implementation on a global scale hinges on the development of photo(electrocatalysts and co-catalysts composed of earth-abundant elements offering visible-light-driven charge separation and surface redox chemistry in high quantum yield, while retaining the chemical and photo-stability typical of titanium dioxide, a ubiquitous oxide semiconductor and performance “benchmark”. The dye-sensitized TiO2 solar cell and multi-junction Si are key “voltage-biasing” components in hybrid photovoltaic/photoelectrochemical (PV/PEC devices that currently lead the field in performance. Prospects and limitations of visible-absorbing particulates, e.g., nanotextured crystalline α-Fe2O3, g-C3N4, and TiO2 sensitized by C/N-based dopants, multilayer composites, and plasmonic metals, are also considered. An interesting trend in water splitting is towards hydrogen peroxide as a solar fuel and value-added green reagent. Fundamental and technical hurdles impeding the advance towards pre-commercial solar fuels demonstration units are considered.

  11. Performance Evaluation of Advanced Ferritic/Martensitic Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    High-chromium(9-12 wt.%) ferritic/martensitic steels are currently being considered as candidate materials for cladding and duct applications in a Gen-IV SFR (sodium-cooled fast reactor) nuclear system because of their higher thermal conductivities and lower expansion coefficients as well as excellent irradiation resistance to void swelling when compared to austenite stainless steels. Since the operation condition in the design of Gen-IV SFR would be envisioned to be harsh from the viewpoints of temperature (≥600 .deg. C) and irradiation dose (≥200 dpa), the primary emphasis is on the fuel cladding materials, i.e. high-Cr ferritic/martensitic steels. The ferritic/martensitic steels for the fuel cladding are commonly used in a 'normalized and tempered' condition. This heat treatment involves a solutionizing treatment (austenitizing) that produces austenite and dissolves the M23C6 carbides and MX carbonitrides, followed by an air cooling that transforms the austenite to martensite. Precipitation sequence during a long-term creep exposure is strongly influenced by the distribution of those in the as heat treated condition of the steels. Their creep strength has been improved by their martensitic lath structure, the precipitation strengthening effects of M23C6 carbides and MX carbonitrides and the solid solution strengthening effects of Mo and W in the matrix. Especially, the precipitation strengthening effect of MX is important because its coarsening rate is small and a fine particle size is maintained for a long-term creep exposure. Z-phase formation from MX-type precipitates has been proposed as a degradation mechanism for a long-term creep regime. The ferritic/martensitic steels should need to improve their performance to be utilized in the high burn-up fuel cladding. For this purpose, KAERI has been developing advanced ferritic/martensitic steels since 2007. This study includes some performance evaluation results of the mechanical and microstructural

  12. Advanced thermally stable jet fuels. Technical progress report, August 1992--October 1992

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Walsh, P.M.; Coleman, M.M.; Bortiatynski, J.; Burgess, C.; Dutta, R.; Gergova, K.; Lai, W.C.; Li, J.; McKinney, D.; Parfitt, D.; Peng, Y.; Sanghani, P.; Yoon, E.

    1993-02-01

    The Penn State program in advanced thermally stable coal-based jet fuels has five borad objectives: (1) development of mechanisms of degradation and solids formation; (2) quantitative measurement of growth of sub-micrometer and miocrometer-sized particles suspended in fuels during thermal stressing; (3) characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and (5) assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Pyrolysis of four isomers of butylbenzene was investigated in static microautoclave reactors at 450{degrees}C under 0.69 MPa of UHP N{sub 2}. Thee rates of disappearance of substrates were found to depend upon the bonding energy of C{alpha}-C{beta} bond in the side chain in the initial period of pyrolysis reactions. Possible catalytic effects of metal surfaces on thermal degradation and deposit formation at temperatures >400{degrees}C have been studied. Carbon deposition depends on the composition of the metal surfaces, and also depends on the chemical compositions of the reactants. Thermal stressing of JP-8 was conducted in the presence of alumina, carbonaceous deposits recovered from earlier stressing experiments, activated carbon, carbon black, and graphite. The addition of different solid carbons during thermal stressing leads to different reaction mechanisms. {sup 13}C NMR spectroscopy, along with {sup 13}C-labeling techniques, have been used to examine the thermal stability of a jet fuel sample mixed with 5% benzyl alcohol. Several heterometallic complexes consisting of two transition metals and sulfur in a single molecule were synthesized and tested as precursors of bimetallic dispersed catalysts for liquefaction of a Montana subbituminous and Pittsburgh No. 8 bituminous coals.

  13. Advanced in fuel channel gauging tool - instrumenting a SLAR tool for dual purpose

    International Nuclear Information System (INIS)

    This paper describes the latest inspection technology to be implemented on a SLARette tool. In 2002, a gauging module was developed and qualified to replace the SLARette tool's blister module. This gauging module had five ultrasonic transducers for diameter creeping and wall thickness measurements. The results of the 2002 SLARette campaign were excellent; the data obtained came within ten microns of that collected with a CANDE tool in 2003. Gentilly-2 decided to continue developing gauging techniques and apparatus to be mounted on a SLARette tool. The new front-end module incorporates both sag measurement and revolutionary pressure tube (PT)/calandria tube (CT) gap modules. Advancements were made along several lines: (1) selection of a radiation-resistant sag module, (2) development of a sag simulator, (3) improvement of AECL gap measurement technology and finally, (4) design of a front-end encompassing module with motorized lift-off capability. This front-end module is only 19 centimeters long and is capable of performing all of the gauging measurements required for fuel channel life-cycle management. This paper will detail the development efforts of Hydro-Quebec, IREQ and AECL in improving fuel channel gauging technology, as well as the implementation and field results of the 2005 Gentilly-2 inspection campaign. (author)

  14. Advanced fuel cell development. Progress Report, April-June 1980. [LiAlO/sub 2/

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R.D.; Arons, R.M.; Dusek, J.T.; Fraioli, A.V.; Kucera, G.H.; Poeppel, R.B.; Sim, J.W.; Smith, J.L.

    1980-11-01

    Advanced fuel cell research and development activities at Argonne National Laboratory (ANL) during the period April-June 1980 are described. These efforts have been directed toward understanding and improving components of molten carbonate fuel cells and have included operation of a 10-cm square cell. Studies have continued on the development of electrolyte structures (LiAlO/sub 2/ and Li/sub 2/CO/sub 3/-K/sub 2/CO/sub 3/). This effort is being concentrated on the preparation of sintered LiAl0/sub 2/ as electrolyte support. Tape casting is presently under investigation as a method for producing green bodies to be sintered; this technique may be an improvement over cold pressing, which was used in the past to produce green bodies. The transition temperature for the ..beta..- to ..gamma..-LiAlO/sub 2/ allotropic transformation is being determined using differential thermal analysis. Work is continuing on the development of preoxidized, prelithiated NiO cathodes. Two techniques, one of which is simpler than the other, have been developed to fabricate plates of Li/sub 0/ /sub 05/Ni/sub 0/ /sub 95/O. In addition, electroless nickel plating is being investigated as a means of providing corrosion protection to structural hardware. To improve its cell testing capability, ANL has constructed a device for improved resistance measurements by the current-interruption technique.

  15. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  16. TYPE A FISSILE PACKAGING FOR AIR TRANSPORT PROJECT OVERVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Eberl, K.; Blanton, P.

    2013-10-11

    This paper presents the project status of the Model 9980, a new Type A fissile packaging for use in air transport. The Savannah River National Laboratory (SRNL) developed this new packaging to be a light weight (<150-lb), drum-style package and prepared a Safety Analysis for Packaging (SARP) for submission to the DOE/EM. The package design incorporates unique features and engineered materials specifically designed to minimize packaging weight and to be in compliance with 10CFR71 requirements. Prototypes were fabricated and tested to evaluate the design when subjected to Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). An overview of the design details, results of the regulatory testing, and lessons learned from the prototype fabrication for the 9980 will be presented.

  17. Examining the stability of thermally fissile Th and U isotopes

    Science.gov (United States)

    Kumar, Bharat; Biswal, S. K.; Singh, S. K.; Patra, S. K.

    2015-11-01

    The properties of recently predicted thermally fissile Th and U isotopes are studied within the framework of the relativistic mean-field approach using the axially deformed basis. We calculate the ground, first intrinsic excited state for highly neutron-rich thorium and uranium isotopes. The possible modes of decay such as α decay and β decay are analyzed. We found that neutron-rich isotopes are stable against α decay, however, they are very unstable against β decay. The lifetime of these nuclei is predicted to be tens of seconds against β decay. If these nuclei are utilized before their decay time, a lot of energy can be produced with the help of multifragmentation fission. Also, these nuclei have great implications from the astrophysical point of view. In some cases, we found that the isomeric states with energy range from 2 to 3 MeV and three maxima in the potential energy surface of Th-230228 and U-234228 isotopes.

  18. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. B. Grover

    2007-05-01

    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-isotropic (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. The 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.1,2 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 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 final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  19. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  20. AREVA NP's advanced Thermal Hydraulic Methods for Reactor Core and Fuel Assembly Design

    International Nuclear Information System (INIS)

    AREVA NP, two converged sub-channel codes have been defined: the homogenous equilibrium model (HEM) code COBRA-FLX and the multi fluid field code F-COBRA-TF. Apart from the sub-channel codes and some smaller specialized codes computational fluid dynamic (CFD) codes are the second important pillar of the AREVA TH code strategy. In the last decade big improvements in the available codes were made and the computing power increased dramatically. Consequently CFD became a reliable and robust tool; thanks to the increased computing power the size of the efficiently calculable models became large enough to be interesting for TH application in fuel assemblies. The main potential of CFD originates from the fact that CFD can predict TH quantities directly, based on the geometric information stored in a computer aided design (CAD) file for mechanic design, the tabulated fluid properties and the desired operating parameters. Hence CFD can be seen as a tool which can be used to perform virtual TH experiments. But unlike experiments where often the access is limited to few TH quantities, CFD provides the comprehensive local TH information and valuable insight into length scales smaller than sub-channels cross sections. Thus, CFD cannot only be used to directly determine the interesting quantities, but also to complement experiments and sub-channel code analysis as well as to support further development of sub-channel codes. AREVA NP's TH methods and codes development strategy follows thus two main streams: 1. Updating and improving the sub-channel codes in order to meet the advanced customer and licensing requirements like improved physical modeling, more detailed information, more flexibility, etc. The recent developments cover the following domains: a. Improved/ Advanced Physics; b. Improved Coding/ Advanced Algorithm. Objective: faster code allowing to perform more calculations or to calculate large models (Pin-by-Pin full core calculations steady state and transient); c

  1. ACSEPT a European project for a new step in the future demonstration of advanced fuel processing

    Energy Technology Data Exchange (ETDEWEB)

    Bourg, S.; Hill, C. [CEA, DRCP - Bat 181, CEA Marcoule, BP17171, 30207 Bagnols/Ceze (France); Caravaca, C.; Espartero, A. [CIEMAT, Avda. Complutense, 22 - 28040 Madrid (Spain); Rhodes, C.; Taylor, R.; Harrison, M. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom); EKBERG, C. [Chalmers tekniska hoegskola, Institutionen foer kemi- och bioteknik, Aemnesomraadets namn, 412 96 Goeteborg (Sweden); GEIST, A. [Forschungszentrum Karlsruhe, Institut fuer Nukleare Entsorgungstechnik, P.O.B. 3640, D-76021 Karlsruhe (Germany); Modolo, G. [Forschungszentrum Juelich - FZJ, D-52425 Juelich (Germany); Cassayre, L. [CNRS, Laboratoire de Genie Chimique, Toulouse (France); Malmbeck, R. [JRC-ITU, Karlsruhe (Germany); De Angelis, G. [ENEA, Casaccia, Rome (Italy); Bouvet, S. [Rio Tinto Alcan, Centre de Recherche de Voreppe, Voreppe (France); Klaassen, F. [NRG, PO Box 25, NL-1755 ZG Petten (Netherlands)

    2010-07-01

    For more than fifteen years, a European scientific community has joined its effort to develop and optimise processes for the partitioning of actinides from fission products. In an international context of 'nuclear renaissance', the upcoming of a new generation of nuclear reactor (Gen IV) will require the development of associated advanced closed fuel cycles which answer the needs of a sustainable nuclear energy: the minimization of the production of long lived radioactive waste but also the optimization of the use of natural resources with an increased resistance to proliferation. Actually, Partitioning and Transmutation (P and T), associated to a multi-recycling of all transuranics (TRUs), should play a key role in the development of this sustainable nuclear energy. By joining together 34 Partners coming from European universities, nuclear research bodies and major industrial players in a multidisciplinary consortium, the FP7 EURATOM-Fission Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), started in 2008 for four year duration, provides the sound basis and fundamental improvements for future demonstrations of fuel treatment in strong connection with fuel fabrication techniques. Consistently with potentially viable recycling strategies, ACSEPT therefore provides a structured R and D framework to develop chemical separation processes compatible with fuel fabrication techniques, with a view to their future demonstration at the pilot level. ACSEPT is organized into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimize and select the most promising ones dedicated either to actinide partitioning or to group actinide separation. (ii) Concerning high temperature pyrochemical separation processes, ACSEPT focuses on the enhancement of the two reference cores of process selected within previous projects. R and D efforts are now devoted to key scientific and technical

  2. Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

    International Nuclear Information System (INIS)

    The current thermal power reactors use less than 1% of the energy contained in uranium. Long term perspectives aiming at a better economical extraction of the potential supplied by uranium motivated the development of new reactor types and, of course, new fuel concepts. Most of them dated from the sixties including liquid metal cooled fast (FR) and high temperature gas cooled (HTGR) reactors. Unfortunately, these impulses slowed down during the last twenty years; nuclear energy had to face political and consensus problems, in particular in the United States of America and in Europe, resulting from the consequences of the TMI and Chernobyl accidents. Good economical results obtained by the thermal power reactors also contributed to this process. During the last twenty years mainly France, India, Japan and the Russian Federation have maintained a relatively high level of technological development with appropriate financial items, in particular, in fuel research for the above mentioned reactor types. China and South Africa are now progressing in development of FR/HTGR and HTGR technologies, respectively. The purpose of this report is not only to summarise knowledge accumulated in the fuel research since the beginning of the sixties. This subject has been well covered in literature up to the end of the eighties. This report rather concentrates on the 'advanced fuels 'for the current different types of reactors including metallic, carbide and nitride fuels for fast reactors, so-called 'cold' fuels and fuels to burn excessive ex-weapons plutonium in thermal power reactors, alternative fuels for small size and research reactors. Emphasis has been put on the aspects of fabrication and irradiation behaviour of these fuels; available basic data concerning essential properties that help to understand the phenomena have been mentioned as well. This report brings complementary information to the earlier published monographs and concerns developments carried out after the early

  3. Advanced Fuel Cycle Initiative AFC-1D, AFC-1G and AFC-1H End of FY-06 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Advanced Fuel Cycle Initiative AFC-1D, AFC-1G and

    2006-09-01

    The U. S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposition and the long-term radiotoxity and heat load of high-level waste sent to a geologic repository. The AFC-1 irradiation experiments on transmutation fuels are expected to provide irradiation performance data on non-fertile and low-fertile fuel forms specifically, irradiation growth and swelling, helium production, fission gas release, fission product and fuel constituent migration, fuel phase equilibria, and fuel-cladding chemical interaction. Contained in this report are the to-date physics evaluations performed on three of the AFC-1 experiments; AFC-1D, AFC-1G and AFC-1H. The AFC-1D irradiation experiment consists of metallic non-fertile fuel compositions with minor actinides for potential use in accelerator driven systems and AFC-1G and AFC-1H irradiation experiments are part of the fast neutron reactor fuel development effort. The metallic fuel experiments and nitride experiment are high burnup analogs to previously irradiated experiments and are to be irradiated to = 40 at.% burnup and = 25 at.% burnup, respectively. Based on the results of the physics evaluations it has been determined that the AFC-1D experiment will remain in the ATR for approximately 4 additional cycles, the AFC-1G experiment for an additional 4-5 cycles, and the AFC-1H experiment for approximately 8 additional cycles, in order to reach the desired programmatic burnup. The specific irradiation schedule for these tests will be determined based on future physics evaluations and all results will be documented in subsequent reports.

  4. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

    1990-08-01

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs.

  5. Enterprise SRS: Leveraging Ongoing Operations to Advance Nuclear Fuel Cycle Programs - 12579

    International Nuclear Information System (INIS)

    The international leadership in nuclear technology development and deployment long held by the United States has eroded due to the lack of clear national strategies for advanced reactor fuel cycle concepts and for nuclear materials management, as well as to the recent policy decision that halts work on the nuclear fuel repository at Yucca Mountain. Although no national consensus on strategy has yet been reached, a number of recent high-profile reviews and workshops have clearly highlighted a national need for robust research, development and deployment (RD and D) programs in key areas of nuclear technology, especially nuclear separations science and engineering. Collectively, these reviews and workshops provide a picture of the nuclear separations mission needs for three major program offices: Department of Energy Office of-Environmental Management), DOE Office of Nuclear Energy), and the National Nuclear Security Administration (NNSA). While the individual program needs differ significantly in detail and timing, they share common needs in two critical areas of RD and D: - The need for access to and use of multi-purpose engineering-scale demonstration test facilities that can support testing with radioactive material, and - The need for collaborative research enterprises that encompass government research organizations (i.e., national laboratories), commercial industry and the academic community. Such collaborative enterprises effectively integrate theory and modeling with the actual experimental work at all scales, as well as strengthen the technical foundation for research in critical areas. The arguments for engineering-scale collaborative research facilities are compelling. Processing history has shown that test programs and demonstrations conducted with actual nuclear materials are essential to program success. It is widely recognized, however, that such facilities are expensive to build and maintain; creating an imposing, if not prohibitive, financial burden

  6. Design and fabrication of an advanced TRISO fuel with ZrC coating

    Energy Technology Data Exchange (ETDEWEB)

    Porter, Ian E., E-mail: porteri@email.sc.edu [University of South Carolina, Mechanical Engineering Department, 300 Main Street, Columbia, SC 29208, United Sates (United States); Knight, Travis W., E-mail: knighttw@cec.sc.edu [University of South Carolina, Mechanical Engineering Department, 300 Main Street, Columbia, SC 29208, United Sates (United States); Dulude, Michael C., E-mail: dulude@email.sc.edu [University of South Carolina, Mechanical Engineering Department, 300 Main Street, Columbia, SC 29208, United Sates (United States); Roberts, Elwyn, E-mail: robertse@cec.sc.edu [University of South Carolina, Mechanical Engineering Department, 300 Main Street, Columbia, SC 29208, United Sates (United States); Hobbs, Jim, E-mail: JSHobbs@nuclearfuelservices.com [Nuclear Fuel Services, Inc., 1205 Banner Hill Road, Erwin, TN 37650 (United States)

    2013-06-15

    Highlights: • Zirconium carbide was deposited on surrogate zirconia and UO{sub 2} kernels. • Deposition rates were found to be dependent on temperature and gas concentration. • Calcining and sintering parameters were optimized to reduce cracking in UO{sub 2} kernel production. -- Abstract: Very high temperature reactors (VHTRs) are expected to achieve coolant outlet temperatures up to 1000 °C, allowing for increased plant efficiency as well as the ability to use the process heat for hydrogen production and various uses in the process chemical industry. The feasibility of using VHTRs as part of the next generation of nuclear reactors greatly depends on the reliability of tri-structural isotropic (TRISO) fuel particles to retain both gaseous and metallic fission products created in irradiated uranium dioxide (UO{sub 2}). This work sought the deposition parameters necessary to produce an additional zirconium carbide (ZrC) layer used in advanced coated particle fuels. The additional ZrC layer will act as an oxygen getter to prevent typical TRISO failure mechanisms including over pressurization of the particle and kernel migration of the kernel within the particle, also known as the amoeba effect. In this study, ZrC coatings were applied to surrogate zirconia kernels as well as UO{sub 2} kernels using a chemical vapor deposition (CVD) fluidized bed reactor, and the deposition characteristics were analyzed via scanning electron microscopy (SEM) techniques. The ZrC layer was confirmed through X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS). The calcining and sintering of urania kernels for use in these coating experiments is also discussed.

  7. Design and fabrication of an advanced TRISO fuel with ZrC coating

    International Nuclear Information System (INIS)

    Highlights: • Zirconium carbide was deposited on surrogate zirconia and UO2 kernels. • Deposition rates were found to be dependent on temperature and gas concentration. • Calcining and sintering parameters were optimized to reduce cracking in UO2 kernel production. -- Abstract: Very high temperature reactors (VHTRs) are expected to achieve coolant outlet temperatures up to 1000 °C, allowing for increased plant efficiency as well as the ability to use the process heat for hydrogen production and various uses in the process chemical industry. The feasibility of using VHTRs as part of the next generation of nuclear reactors greatly depends on the reliability of tri-structural isotropic (TRISO) fuel particles to retain both gaseous and metallic fission products created in irradiated uranium dioxide (UO2). This work sought the deposition parameters necessary to produce an additional zirconium carbide (ZrC) layer used in advanced coated particle fuels. The additional ZrC layer will act as an oxygen getter to prevent typical TRISO failure mechanisms including over pressurization of the particle and kernel migration of the kernel within the particle, also known as the amoeba effect. In this study, ZrC coatings were applied to surrogate zirconia kernels as well as UO2 kernels using a chemical vapor deposition (CVD) fluidized bed reactor, and the deposition characteristics were analyzed via scanning electron microscopy (SEM) techniques. The ZrC layer was confirmed through X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS). The calcining and sintering of urania kernels for use in these coating experiments is also discussed

  8. The environmental impacts of Korean advanced nuclear fuel cycle KIEP-21 and disposal concepts

    International Nuclear Information System (INIS)

    We have performed a performance assessment to investigate effects of waste forms and repository designs by comparing the case of direct disposal of used PWR fuel in the Korean Reference Repository System (KRS) concept with the case of Advanced Korean Reference Disposal System (A-KRS) repository containing ILW and HLW from the KIEP-21 system. Numerical evaluations have been made for release rates of actinide and fission product isotopes at the boundaries of the engineered barrier system (EBS) and the natural barrier system (NBS) by the TTB code developed at UC Berkeley. Results show that in both cases, most actinides and their daughters remain as precipitates in the EBS because of their assumed low solubilities. The radionuclides that reach the 1 000-m location in NBS are fission products, 129I, 79Se and 36Cl. They have high solubilities and weak or no sorption with the EBS materials or with the host rock, and are released congruently with waste form alteration. In case of direct disposal, a contribution of 2% of iodine is assumed to be accumulated in the gap between the cladding and fuel pellets released after failure of the waste package and cladding dominates the total release rate. With increase in the waste form alteration time, the peak value of total release rate decreases proportionally because the dominant radionuclides are fission product isotopes, which are released from waste forms congruently with waste form dissolution. It has been shown by PHREEQC simulation that actinide solubilities can be significantly affected by pore water chemistry determined by the evolving EBS materials, waste forms and compositions of groundwater from the far field. (authors)

  9. Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

    Science.gov (United States)

    Ribeiro, Celso

    2015-11-01

    The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

  10. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    Energy Technology Data Exchange (ETDEWEB)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  11. Electricity and fluid fuels from biomass and coal using advanced technologies: a cost comparison for developing country applications

    International Nuclear Information System (INIS)

    Recent analyses of alternative global energy supply strategies, such as the forthcoming report of the Intergovernmental Panel on Climate Change (IPCC), to be published in 1996, have drawn attention to the possibility that biomass modernized with advanced technologies could play an important role in meeting global energy needs in the next century. This paper discusses two promising classes of advanced technologies that offer the potential for providing modem energy carriers (electricity and fluid fuels) from biomass at competitive costs within one or two decades. These technologies offer significantly more efficient use of land than currently commercial technologies for producing electricity and fluid fuels from biomass, as well as substantially improved energy balances. Electricity is Rely to be the first large market for modernized biomass, but the potential market for fluid fuel production is likely to be much larger. As coal is likely to present a more serious competitive challenge to biomass in the long run, we present an economic comparison with coal-based electricity and fluid fuels. A meaningful economic comparison between coal and biomass is possible because these feedstocks are sufficiently alike in their physical characteristics that similar conversion technologies may well be used for producing electricity and fluid fuels from them. When similar conversion technologies are used for both feedstocks, the relative costs of electricity or fluid fuels will be determined by the distinguishing technical characteristics of the feedstocks (sulphur content, moisture content and reactivity) and by the relative feedstock prices. Electric power generation from biomass and coal are compared here using an advanced integrated gasifier/gas turbine cycle that offers the potential for achieving high efficiency, low unit capital cost and low local pollutant emissions: the steam-injected gas turbine coupled to an air-blown gasifier. For both feedstocks, generation costs are

  12. Evaluation and development of advanced nuclear materials: IAEA activities

    International Nuclear Information System (INIS)

    Economical, environmental and non-proliferation issues associated with sustainable development of nuclear power bring about a need for optimization of fuel cycles and implementation of advanced nuclear systems. While a number of physical and design concepts are available for innovative reactors, the absence of reliable materials able to sustain new challenging irradiation conditions represents the real bottle-neck for practical implementation of these promising ideas. Materials performance and integrity are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy production incorporating fuel recycling and waste transmutation systems. These systems will feature high thermal operational efficiency, improved utilization of resources (both fissile and fertile materials) and reduced production of nuclear waste. They will require development, qualification and deployment of new and advanced fuel and structural materials with improved mechanical and chemical properties combined with high radiation and corrosion resistance. The extensive, diverse, and expensive efforts toward the development of these materials can be more effectively organized within international collaborative programmes with wide participation of research, design and engineering communities. IAEA carries out a number of international projects supporting interested Member States with the use of available IAEA program implementation tools (Coordinated Research Projects, Technical Meetings, Expert Reviews, etc). The presentation summarizes the activities targeting material developments for advanced nuclear systems, with particular emphasis on fast reactors, which are the focal topics of IAEA Coordinated Research Projects 'Accelerator Simulation and Theoretical Modelling of Radiation Effects' (on-going), 'Benchmarking of Structural Materials Pre-Selected for Advanced Nuclear Reactors', 'Examination of advanced fast reactor fuel and core

  13. R&D plan for immobilization technologies: fissile materials disposition program. Revision 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, H.F.; Armantrout, G.A.

    1996-09-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long- term fissile material management options, the Department of Energy`s Fissile Materials Disposition Program (FMDP) is conducting studies of options for the storage and disposition of surplus plutonium (Pu). One set of alternatives for disposition involve immobilization. The immobilization alternatives provide for fixing surplus fissile materials in a host matrix in order to create a solid disposal form that is nuclear criticality-safe, proliferation-resistant and environmentally acceptable for long-term storage or disposal.

  14. Process Modeling Phase I Summary Report for the Advanced Gas Reactor Fuel Development and Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

    Pannala, Sreekanth [ORNL; Daw, C Stuart [ORNL; Boyalakuntla, Dhanunjay S [ORNL; FINNEY, Charles E A [ORNL

    2006-09-01

    This report summarizes the results of preliminary work at Oak Ridge National Laboratory (ORNL) to demonstrate application of computational fluid dynamics modeling to the scale-up of a Fluidized Bed Chemical Vapor Deposition (FBCVD) process for nuclear fuels coating. Specifically, this work, referred to as Modeling Scale-Up Phase I, was conducted between January 1, 2006 and March 31, 2006 in support of the Advanced Gas Reactor (AGR) Program. The objective was to develop, demonstrate and "freeze" a version of ORNL's computational model of the TRI ISOtropic (TRISO) fuel-particle coating process that can be specifically used to assist coater scale-up activities as part of the production of AGR-2 fuel. The results in this report are intended to serve as input for making decisions about initiating additional FBCVD modeling work (referred to as Modeling Scale-Up Phase II) in support of AGR-2. The main computational tool used to implement the model is the general-purpose multiphase fluid-dynamics computer code known as MFIX (Multiphase Flow with Interphase eXchanges), which is documented in detail on the DOE-sponsored website http://www.mfix.org. Additional computational tools are also being developed by ORNL for post-processing MFIX output to efficiently summarize the important information generated by the coater simulations. The summarized information includes quantitative spatial and temporal measures (referred to as discriminating characteristics, or DCs) by which different coater designs and operating conditions can be compared and correlated with trends in product quality. The ORNL FBCVD modeling work is being conducted in conjunction with experimental coater studies at ORNL with natural uranium CO (NUCO) and surrogate fuel kernels. Data are also being obtained from ambient-temperature, spouted-bed characterization experiments at the University of Tennessee and theoretical studies of carbon and silicon carbide chemical vapor deposition kinetics at Iowa State

  15. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  16. Status on Establishing the Feasibility of Lead Slowing Down Spectroscopy for Direct Measurement of Plutonium in Used Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.; Gavron, Victor A.; Devlin, M.; Haight, R. C.; O' Donnell, J. M.; Danon, Yaron; Weltz, Adam; Bonebrake, Eric; Imel, G. R.; Harris, Jason; Beller, Dennis; Hatchett, D.; Droessler, J.

    2012-08-30

    Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to study the feasibility of Lead Slowing Down Spectroscopy. This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today’s confirmatory assay methods. This paper will present efforts on the development of time-spectral analysis algorithms, fast neutron detector advances, and validation and testing measurements.

  17. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles, September 12-14, 2005

    International Nuclear Information System (INIS)

    The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear

  18. A concept of prospective sodium fast reactor with ductless fuel subassemblies in the core

    Energy Technology Data Exchange (ETDEWEB)

    Sedov, A.A.; Alekseev, P.N.; Fomichenko, P.A.; Ponomarev-Stepnoy, N.N.; Proshkin, A.A.; Ponomarev, A.S.; Stukalov, V.A. [Russian Research Center, Kurchatov Institute, Moscow (Russian Federation)

    2007-07-01

    The Kurchatov Institute studies the concept of a sodium fast reactor (SFR) with advanced core design, which is based on the following principle technique solutions: -) application of ductless fuel subassemblies with wide lattice of fuel rods of increased diameter and spaced by grids; -) the usage of dense U-Pu ceramic fuel and low-nickel steels, and -) application of cluster-type control and protection system. Preconceptual studies have shown, that SFR with advanced core design is 3 times more effective in the fuel consumption than project BN-800 reactor due to better neutron balance in the core and CBR (core breeding ratio) {approx} 1, provides getting quite high burn-up of the core fuel (Bmax {approx} 15-20 % of heavy atoms), increases fuel life up to 7-8 years at specific loading of fissile nuclides in the core less than 5 t/GW, decreases electricity demand for pumping the primary coolant (due to low hydraulic resistance of the core) and has bigger safety potential in accidents than the core with traditional liquid metal fast reactor design (due to low core reactivity margin, high level of natural circulation and subassemblies hydraulic interaction). In the paper the main results of preconceptual feasibility study of SFR with advanced core design are presented and discussed with a focus on technique and economic aspects. Some of characteristic features of core neutron physics, thermal hydraulics and fuel rod thermal mechanics behavior are displayed and discussed as well. (authors)

  19. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  20. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    Science.gov (United States)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  1. An advanced model for grain face diffusion transport in irradiated UO{sub 2} fuel. Part 1: Model formulation

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S., E-mail: vms@ibrae.ac.r [Nuclear Safety Institute (IBRAE), Russian Academy of Sciences, 52, B. Tulskaya, Moscow 115191 (Russian Federation); Tarasov, V.I. [Nuclear Safety Institute (IBRAE), Russian Academy of Sciences, 52, B. Tulskaya, Moscow 115191 (Russian Federation)

    2009-07-01

    An advanced model for the grain face transport of gas atoms, self-consistently taking into consideration the effects of atom diffusion over the grain surface, their trapping by and irradiation induced resolution from intergranular bubbles is presented. The model allows prediction of a noticeable gas release from UO{sub 2} fuel without visible interlinkage of grain face bubbles, i.e. at very low grain face coverage, below the critical value manifested by formation of bubble channels on grain faces interconnected with open porosity, in accordance with experimental observations of UO{sub 2} and MOX fuel behaviour under various irradiation conditions.

  2. Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

    International Nuclear Information System (INIS)

    The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO2-UO2 fissile materials and UO2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange

  3. Implications of Results from the Advanced Gas Reactor Fuel Development and Qualification Program on Licensing of Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2001-10-01

    The high level of safety of modular high temperature gas-cooled reactor (HTGR) designs is achieved by passively maintaining core temperatures below fission product release thresholds under all envisioned accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to other reactor types but is predicated on exceptionally high coated-particle fuel fabrication quality and excellent fuel performance under normal operation and accident conditions. The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program decided to qualify for uranium oxide/uranium carbide (UCO) TRISO coated-particle fuel in an operating envelope that would bound both pebble bed and prismatic modular HTGR options. By using a mixture of uranium oxide and uranium carbide, the kernel composition is engineered to minimize CO formation and fuel kernel migration, which is key to maintain to fuel integrity at the higher burnups, temperatures, and temperature gradients anticipated in prismatic HTGRs. Fuel fabrication conducted at both laboratory and engineering scale has demonstrated the ability to fabricate high quality UCO TRISO fuel with very low defects. The first irradiation (AGR 1) exposed about 300,000 TRISO fuel particles to a peak burnup of 19.6% FIMA, a peak fast-neutron fluence of about 4.3 × 1025 n/m2, and a maximum time-averaged fuel temperature of about 1,200°C without a single particle failure. The very low release of key metallic fission products (except silver) measured in post-irradiation examination (PIE) confirms the excellent performance measured under irradiation. Very low releases have been measured in accident simulation heatup testing (''safety testing'') after hundreds of hours at 1600 and 1700°C and no particle failures (no noble gas release measured) have been observed. Even after hundreds of hours at 1800°C, the releases are

  4. LiNiFe-based layered structure oxide and composite for advanced single layer fuel cells

    Science.gov (United States)

    Zhu, Bin; Fan, Liangdong; Deng, Hui; He, Yunjune; Afzal, Muhammad; Dong, Wenjing; Yaqub, Azra; Janjua, Naveed K.

    2016-06-01

    A layered structure metal oxide, LiNi0.1Fe0.90O2-δ (LNF), is explored for the advanced single layer fuel cells (SLFCs). The temperature dependent impedance profiles and concentration cells (hydrogen concentration, oxygen concentration, and H2/air atmospheres) tests prove LNF to be an intrinsically electronic conductor in air while mixed electronic and proton conductor in H2/air environment. SLFCs constructed by pure LNF materials show significant short circuiting reflected by a low device OCV and power output (175 mW cm-2 at 500 °C) due to high intrinsic electronic conduction. The power output is improved up to 640 and 760 mW cm-2, respectively at 500 and 550 °C by compositing LNF with ion conducting material, e.g., samarium doped ceria (SDC), to balance the electronic and ionic conductivity; both reached at 0.1 S cm-1 level. Such an SLFC gives super-performance and simplicity over the conventional 3-layer (anode, electrolyte and cathode) FCs, suggesting strong scientific and commercial impacts.

  5. Advanced DC-DC converter for power conditioning in hydrogen fuel cell systems

    Energy Technology Data Exchange (ETDEWEB)

    Kovacevic, G.; Tenconi, A.; Bojoi, R. [Department of Electrical Engineering, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2008-06-15

    The fuel cell (FC) generators can produce electric energy directly from hydrogen and oxygen. The DC voltage generated by FC is generally low amplitude and it is not constant, depending on the operating conditions. Furthermore, FC systems have dynamic response that is slower than the transient responses typically requested by the load. For this reason, in many applications the FC generators must be interfaced with other energy/power sources by means of an electronic power converter. An advanced full-bridge (FB) DC-DC converter, which effectively achieves zero-voltage switching and zero-current switching (ZVS-ZCS), is proposed for power-conditioning (PC) in hydrogen FC applications. The operation and features of the converter are analyzed and verified by simulations results. The ZVS-ZCS operation is obtained by means of a simple auxiliary circuit. Introduction of the soft-switching operation in PC unit brings improvements not only from the converter efficiency point of view, but also in terms of increased converter power density. Quantitative analysis of hard and soft-switching operating of the proposed converter is also made, bringing in evidence the benefits of soft-switching operation mode. The proposed converter can be a suitable solution for PC in hydrogen FC systems, especially for the medium to high-power applications. (author)

  6. Diamond and Hydrogenated Carbons for Advanced Batteries and Fuel Cells: Fundamental Studies and Applications.

    Energy Technology Data Exchange (ETDEWEB)

    Swain; Greg M.

    2009-04-13

    The original funding under this project number was awarded for a period 12/1999 until 12/2002 under the project title Diamond and Hydrogenated Carbons for Advanced Batteries and Fuel Cells: Fundamental Studies and Applications. The project was extended until 06/2003 at which time a renewal proposal was awarded for a period 06/2003 until 06/2008 under the project title Metal/Diamond Composite Thin-Film Electrodes: New Carbon Supported Catalytic Electrodes. The work under DE-FG02-01ER15120 was initiated about the time the PI moved his research group from the Department of Chemistry at Utah State University to the Department of Chemistry at Michigan State University. This DOE-funded research was focused on (i) understanding structure-function relationships at boron-doped diamond thin-film electrodes, (ii) understanding metal phase formation on diamond thin films and developing electrochemical approaches for producing highly dispersed electrocatalyst particles (e.g., Pt) of small nominal particle size, (iii) studying the electrochemical activity of the electrocatalytic electrodes for hydrogen oxidation and oxygen reduction and (iv) conducting the initial synthesis of high surface area diamond powders and evaluating their electrical and electrochemical properties when mixed with a Teflon binder.

  7. Advanced and flexible genetic algorithms for BWR fuel loading pattern optimization

    International Nuclear Information System (INIS)

    This work proposes advances in the implementation of a flexible genetic algorithm (GA) for fuel loading pattern optimization for Boiling Water Reactors (BWRs). In order to avoid specific implementations of genetic operators and to obtain a more flexible treatment, a binary representation of the solution was implemented; this representation had to take into account that a little change in the genotype must correspond to a little change in the phenotype. An identifier number is assigned to each assembly by means of a Gray Code of 7 bits and the solution (the loading pattern) is represented by a binary chain of 777 bits of length. Another important contribution is the use of a Fitness Function which includes a Heuristic Function and an Objective Function. The Heuristic Function which is defined to give flexibility on the application of a set of positioning rules based on knowledge, and the Objective Function that contains all the parameters which qualify the neutronic and thermal hydraulic performances of each loading pattern. Experimental results illustrating the effectiveness and flexibility of this optimization algorithm are presented and discussed.

  8. Engineering development of advance physical fine coal cleaning for premium fuel applications

    Energy Technology Data Exchange (ETDEWEB)

    Jha, M.C.; Smit, F.J.; Shields, G.L. [AMAX R& D Center/ENTECH Global Inc., Golden, CO (United States)

    1995-11-01

    The objective of this project is to develop the engineering design base for prototype fine coal cleaning plants based on Advanced Column Flotation and Selective Agglomeration processes for premium fuel and near-term applications. Removal of toxic trace elements is also being investigated. The scope of the project includes laboratory research and bench-scale testing of each process on six coals followed by design, construction, and operation of a 2 tons/hour process development unit (PDU). Three coals will be cleaned in tonnage quantity and provided to DOE and its contractors for combustion evaluation. Amax R&D (now a subsidiary of Cyprus Amax Mineral Company) is the prime contractor. Entech Global is managing the project and performing most of the research and development work as an on-site subcontractor. Other participants in the project are Cyprus Amax Coal Company, Arcanum, Bechtel, TIC, University of Kentucky and Virginia Tech. Drs. Keller of Syracuse and Dooher of Adelphi University are consultants.

  9. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Day, Christy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  10. Fissile Material Detection by Differential Die Away Analysis

    Science.gov (United States)

    Shaw, Timothy J.; Strellis, Dan A.; Stevenson, John; Keeley, Doug; Gozani, Tsahi

    2009-03-01

    Detection and interdiction of Special Nuclear Material (SNM) in transportation is one of the most critical security issues facing the United States. Active inspection by inducing fission in fissile nuclear materials, such as 235U and 239Pu, provides several strong and unique signatures that make the detection of concealed nuclear materials technically very feasible. Differential Die-Away Analysis (DDAA) is a very efficient, active neutron-based technique that uses the abundant prompt fission neutrons signature. It benefits from high penetrability of the probing and signature neutrons, high fission cross section, high detection sensitivity, ease of deployment and relatively low cost. DDAA can use any neutron source or energy as long as it can be suitably pulsed. The neutron generator produces pulses of neutrons that are directed into a cargo. As each pulse passes through the cargo, the neutrons are thermalized and absorbed. If SNM is present, the thermalized neutrons create a new source of (fission) neutrons with a distinctive time profile. An efficient laboratory system was designed, fabricated and tested under a US Government DHS DNDO contract. It was shown that a small uranium sample can be detected in a large variety of cargo types and configurations within practical measurement times using commercial compact (d,T) sources. Using stronger sources and wider detector distribution will further cut inspection time. The system can validate or clear alarms from a primary inspection system such as an automated x-ray system.

  11. Examining the stability of thermally fissile Th and U isotopes

    CERN Document Server

    Kumar, Bharat; Singh, S K; Patra, S K

    2015-01-01

    The properties of recently predicted thermally fissile Th and U isotopes are studied within the framework of relativistic mean field (RMF) approach using axially deformed basis. We calculated the ground, first intrinsic excited state and matter density for highly neutron-rich thorium and uranium isotopes. The possible modes of decay like $\\alpha$-decay and $\\beta$-decay are analyzed. We found that the neutron-rich isotopes are stable against $\\alpha$-decay, however they are very much unstable against $\\beta$-decay. The life time of these nuclei predicted to be tens of second against $\\beta$-decay. If these nuclei utilize before their decay time, a lots of energy can be produced within the help of multi-fragmentation fission. Also, these nuclei have a great implication in astrophysical point of view. The total nucleonic densities distribution are calculated, from which the clusters inside the parent nuclei are determined. %Most of the thorium isotopes are $\\alpha$ emitters, where as some %of them have short ha...

  12. Extensions to Dynamic System Simulation of Fissile Solution Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bernardin, John David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Spernjak, Dusan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-24

    Previous reports have documented the results of applying dynamic system simulation (DSS) techniques to model a variety of fissile solution systems. The SUPO (Super Power) aqueous homogeneous reactor (AHR) was chosen as the benchmark for comparison of model results to experimental data for steadystate operation.1 Subsequently, DSS was applied to additional AHR to verify results obtained for SUPO and extend modeling to prompt critical excursions, ramp reactivity insertions of various magnitudes and rate, and boiling operations in SILENE and KEWB (Kinetic Experiment Water Boiler).2 Additional models for pressurized cores (HRE: Homogeneous Reactor Experiment), annular core geometries, and accelerator-driven subcritical systems (ADAHR) were developed and results reported.3 The focus of each of these models is core dynamics; neutron kinetics, thermal hydraulics, radiolytic gas generation and transport are coupled to examine the time-based evolution of these systems from start-up through transition to steady-state. A common characteristic of these models is the assumption that (a) core cooling system inlet temperature and flow and (b) plenum gas inlet pressure and flow are held constant; no external (to core) component operations that may result in dynamic change to these parameters are considered. This report discusses extension of models to include explicit reference to cooling structures and radiolytic gas handling. The accelerator-driven subcritical generic system model described in References 3 and 4 is used as a basis for this extension.

  13. Update to the Fissile Materials Disposition program SST/SGT transportation estimation

    International Nuclear Information System (INIS)

    This report is an update to ''Fissile Materials Disposition Program SST/SGT Transportation Estimation,'' SAND98-8244, June 1998. The Department of Energy Office of Fissile Materials Disposition requested this update as a basis for providing the public with an updated estimation of the number of transportation loads, load miles, and costs associated with the preferred alternative in the Surplus Plutonium Disposition Final Environmental Impact Statement (EIS)

  14. Ethanol as a fuel for road transportation. Main report; Contribution to IEA Implementing Agreement on Advanced Motor Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Ulrik; Johansen, T.; Schramm, J.

    2009-05-15

    Bioethanol as a motor fuel in the transportation sector, mainly for road transportation, has been subject to many studies and much discussion. Furthermore, the topic involves not only the application and engine technical aspects, but also the understanding of the entire life cycle of the fuel, well-to-wheels, including economical, environmental, and social aspects. It is not, however, the aim of this report to assess every single one of these aspects. The present report aims to address the technical potential and problems as well as the central issues related to the general application of bioethanol as an energy carrier in the near future. In discussions of the advantages and drawbacks of ethanol, the type of application is important. Generalization is not possible, because ethanol can be used in many forms. Furthermore, a wide range of ethanol/gasoline blends has not yet been investigated sufficiently. The most favorable type of application is determined by infrastructural factors, especially vehicle fleet configuration. From a technical point of view, optimal usage involves a high degree of water content in the ethanol, and this excludes low-percentage-ethanol fuels. The benefits seem strongly related to the amount of ethanol in a given blend, that is, the more the better. Both engine efficiencies and emissions improve with more ethanol in the fuel. Wet ethanol constitutes an even cleaner fuel in both the production and application phases. In summary, ethanol application has many possibilities, but with each type of application comes a set of challenges. Nevertheless, technical solutions for each challenge are available. (ln)

  15. Recent advances in microbial production of fuels and chemicals using tools and strategies of systems metabolic engineering.

    Science.gov (United States)

    Cho, Changhee; Choi, So Young; Luo, Zi Wei; Lee, Sang Yup

    2015-11-15

    The advent of various systems metabolic engineering tools and strategies has enabled more sophisticated engineering of microorganisms for the production of industrially useful fuels and chemicals. Advances in systems metabolic engineering have been made in overproducing natural chemicals and producing novel non-natural chemicals. In this paper, we review the tools and strategies of systems metabolic engineering employed for the development of microorganisms for the production of various industrially useful chemicals belonging to fuels, building block chemicals, and specialty chemicals, in particular focusing on those reported in the last three years. It was aimed at providing the current landscape of systems metabolic engineering and suggesting directions to address future challenges towards successfully establishing processes for the bio-based production of fuels and chemicals from renewable resources.

  16. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  17. 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

  18. High-resolution computed tomography for flaw detection in advanced thin-layer ceramics for fuel cells

    International Nuclear Information System (INIS)

    Advanced monolithic solid oxide fuel cells are being assembled from thin (∼50 μm) tape-cast ceramic layers with highly tailored mechanical properties. The layers need to be free of cracks and nonbonds. A high-resolution computed tomography system with a Ir-192 source was investigated as a tool for detecting cracks and nonbonds. Results suggest that channel sizes, including internal channels, can be determined but small in-plane cracks cannot be detected

  19. Methods and apparatuses for the development of microstructured nuclear fuels

    Science.gov (United States)

    Jarvinen, Gordon D.; Carroll, David W.; Devlin, David J.

    2009-04-21

    Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

  20. Hydrogen as a fuel for today and tomorrow: expectations for advanced hydrogen storage materials/systems research.

    Science.gov (United States)

    Hirose, Katsuhiko

    2011-01-01

    History shows that the evolution of vehicles is promoted by several environmental restraints very similar to the evolution of life. The latest environmental strain is sustainability. Transport vehicles are now facing again the need to advance to use sustainable fuels such as hydrogen. Hydrogen fuel cell vehicles are being prepared for commercialization in 2015. Despite intensive research by the world's scientists and engineers and recent advances in our understanding of hydrogen behavior in materials, the only engineering phase technology which will be available for 2015 is high pressure storage. Thus industry has decided to implement the high pressure tank storage system. However the necessity of smart hydrogen storage is not decreasing but rather increasing because high market penetration of hydrogen fuel cell vehicles is expected from around 2025 onward. In order to bring more vehicles onto the market, cheaper and more compact hydrogen storage is inevitable. The year 2025 seems a long way away but considering the field tests and large scale preparation required, there is little time available for research. Finding smart materials within the next 5 years is very important to the success of fuel cells towards a low carbon sustainable world.

  1. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

    Advances in gas centrifuges and development of the atomic vapour laser isotope separation process promise substantial reductions in the cost of enriched uranium. The resulting reduction in LWR fuel costs could seriously erode the economic advantage of CANDU, and in combination with LWR design improvements, shortened construction times and increased operational reliability could allow the LWR to overtake CANDU. CANDU's traditional advantages of neutron economy and high reliability may no longer be sufficient - this is the challenge. The responses include: combining neutron economy and dollar economy by optimizing CANDU for slightly enriched uranium fuel; developing cost-reducing improvements in design, manufacture and construction; and reducing the cost of heavy water. Technology is a renewable resource which must be continually applied to a product for it to remain competitive in the decades to come. Such innovation is a prerequisite to Canada increasing her share of the international market for nuclear power stations. The higher burn-up achievable with enriched fuel in CANDU can reduce the fuel cycle costs by 20 to 40 percent for a likely range of costs for yellowcake and separative work. Alternatively, some of the benefits of a higher fissile content can take the form of a cheaper reactor core containing fewer fuel channels and less heavy water, and needing only a single fuelling machine. An opportunity that is linked to this need to introduce an enriched uranium fuel cycle into CANDU is to build an enrichment business in Canada. This could offer greater value added to our uranium exports, security of supply for enriched CANDUs, technological growth in Canada and new employment opportunities. AECL has a study in progress to define this opportunity

  2. The attractiveness of materials in advanced nuclear fuel cycles for various proliferation and theft scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, Charles G [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Ireland, John R [Los Alamos National Laboratory; Johnson, M W [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Ebbinghaus, Bartley B [LLNL; Sleaford, Brad A [LLNL; Bradley, Keith S [LLNL; Collins, Brian W [PNNL; Smith, Brian W [PNNL; Prichard, Andrew W [PNNL

    2009-01-01

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  3. The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Hase, Kevin R.; Jarvinen, G. D.; Ebbinghaus, B. B.; Sleaford, Brad W.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

    2010-09-01

    This paper is an extension to earlier studies1,2 that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no “silver bullet” has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of "attractiveness levels" that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities.3 The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  4. The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.; Wallace, R.K.; Hase, K.R.; Jarvinen, G.D.; Ireland, J.R.; Johnson, M.W. [Los Alamos National Laboratory, P.O. Box 1663, MS K575, Los Alamos, NM 87545 (United States); Ebbinghaus, B.B.; Sleaford, B.W.; Bradley, K.S. [Lawrence Livermore National Laboratory (United States); Collins, B.A.; Prichard, A.W.; Smith, B.W. [Pacific Northwest National Laboratory (United States)

    2009-06-15

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed. (authors)

  5. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  6. Safety and Regulatory Issues of the Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

    2014-02-01

    Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

  7. Advanced thermally stable jet fuels: Technical progress report, October 1994--December 1994

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Boehman, A.; Coleman, M.M.

    1995-02-01

    There are five tasks within this project on thermally stable coal-based jet fuels. Progress on each of the tasks is described. Task 1, Investigation of the quantitative degradation chemistry of fuels, has 5 subtasks which are described: Literature review on thermal stability of jet fuels; Pyrolytic and catalytic reactions of potential endothermic fuels: cis- and trans-decalin; Use of site specific {sup 13}C-labeling to examine the thermal stressing of 1-phenylhexane: A case study for the determination of reaction kinetics in complex fuel mixtures versus model compound studies; Estimation of critical temperatures of jet fuels; and Surface effects on deposit formation in a flow reactor system. Under Task 2, Investigation of incipient deposition, the subtask reported is Uncertainty analysis on growth and deposition of particles during heating of coal-derived aviation gas turbine fuels; under Task 3, Characterization of solid gums, sediments, and carbonaceous deposits, is subtask, Studies of surface chemistry of PX-21 activated carbon during thermal degradation of jet A-1 fuel and n-dodecane; under Task 4, Coal-based fuel stabilization studies, is subtask, Exploratory screening and development potential of jet fuel thermal stabilizers over 400 C; and under Task 5, Exploratory studies on the direct conversion of coal to high quality jet fuels, are 4 subtasks: Novel approaches to low-severity coal liquefaction and coal/resid co-processing using water and dispersed catalysts; Shape-selective naphthalene hydrogenation for production of thermally stable jet fuels; Design of a batch mode and a continuous mode three-phase reactor system for the liquefaction of coal and upgrading of coal liquids; and Exploratory studies on coal liquids upgrading using mesopores molecular sieve catalysts. 136 refs., 69 figs., 24 tabs.

  8. The Economic, repository and proliferation implications of advanced nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark; Cady, K B

    2011-09-04

    The goal of this project was to compare the effects of recycling actinides using fast burner reactors, with recycle that would be done using inert matrix fuel burned in conventional light water reactors. In the fast reactor option, actinides from both spent light water and fast reactor fuel would be recycled. In the inert matrix fuel option, actinides from spent light water fuel would be recycled, but the spent inert matrix fuel would not be reprocessed. The comparison was done over a limited 100-year time horizon. The economic, repository and proliferation implications of these options all hinge on the composition of isotopic byproducts of power production. We took the perspective that back-end economics would be affected by the cost of spent fuel reprocessing (whether conventional uranium dioxide fuel, or fast reactor fuel), fuel manufacture, and ultimate disposal of high level waste in a Yucca Mountain like geological repository. Central to understanding these costs was determining the overall amount of reprocessing needed to implement a fast burner, or inert matrix fuel, recycle program. The total quantity of high level waste requiring geological disposal (along with its thermal output), and the cost of reprocessing were also analyzed. A major advantage of the inert matrix fuel option is that it could in principle be implemented using the existing fleet of commercial power reactors. A central finding of this project was that recycling actinides using an inert matrix fuel could achieve reductions in overall actinide production that are nearly very close to those that could be achieved by recycling the actinides using a fast burner reactor.

  9. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  10. Evaluation of the Use of Synroc to Solidify the Cesium and Strontium Separations Product from Advanced Aqueous Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Julia Tripp; Vince Maio

    2006-03-01

    This report is a literature evaluation on the Synroc process for determining the potential for application to solidification of the Cs/Sr strip product from advanced aqueous fuel separations activities.

  11. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  12. The role of advanced calculation and simulation tools in the evolution of fuel; El papel de las herramientas avanzadas de calculo y simulacion en la evolucion del combustible

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Reja, C.; Cerracin, A.; Corpa, R.

    2015-07-01

    This article is focused on the role of the advanced calculation/simulation tools on the development of the