Sample records for advanced burner reactors

  1. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

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

    2008-12-16T23:59:59.000Z

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

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

    SciTech Connect (OSTI)

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

    2006-09-15T23:59:59.000Z

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

  3. Core design studies for advanced burner test reactor.

    SciTech Connect (OSTI)

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01T23:59:59.000Z

    The U.S. government announced in February 2006 the Global Nuclear Energy Partnership (GNEP) to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. The advanced burner reactor (ABR) based on a fast spectrum is one of the three major technologies to be demonstrated in GNEP. In FY06, a pre-conceptual design study was performed to develop an advanced burner test reactor (ABTR) that supports development of a prototype full-scale ABR, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR were (1) to demonstrate reactor-based transmutation of transuranics (TRU) as part of an advanced fuel cycle, (2) to qualify the TRU-containing fuels and advanced structural materials needed for a full-scale ABR, (3) to support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. Based on these objectives, core design and fuel cycle studies were performed to develop ABTR core designs, which can accommodate the expected changes of the TRU feed and the conversion ratio. Various option and trade-off studies were performed to determine the appropriate power level and conversion ratio. Both ternary metal alloy (U-TRU-10Zr) and mixed oxide (UO{sub 2}-TRUO{sub 2}) fuel forms have been considered with TRU feeds from weapons-grade plutonium (WG-Pu) and TRU recovered from light water reactor spent fuel (LWR-SF). Reactor performances were evaluated in detail including equilibrium cycle core parameters, mass flow, power distribution, kinetic parameters, reactivity feedback coefficient, reactivity control requirements and shutdown margins, and spent fuel characteristics. Trade-off studies on power level suggested that about 250 MWt is a reasonable compromise to allow a low project cost, at the same time providing a reasonable prototypic irradiation environment for demonstrating TRU-based fuels. Preliminary design studies showed that it is feasible to design the ABTR to accommodate a wide range of conversion ratio (CR) by employing different assembly designs. The TRU enrichments required for various conversion ratios and the irradiation database suggested a phased approach with initial startup using conventional enrichment plutonium-based fuel and gradual transitioning to full core loading of transmutation fuel after its qualification phase (resulting in {approx}0.6 CR). The low CR transmutation fuel tests can be accommodated in the designated test assemblies, and if fully developed, core conversion to low CR fuel can be envisioned. Reference ABTR core designs with a rated power of 250 MWt were developed for ternary metal alloy and mixed oxide fuels based on WG-Pu feed. The reference core contains 54 driver, 6 test fuel, and 3 test material assemblies. For the startup core designs, the calculated TRU conversion ratio is 0.65 for the metal fuel core and 0.64 for the oxide fuel core. Both the metal and oxide cores show good performances. The metal fuel core requires an average TRU enrichment of 18.8% and yields a reactivity swing of 1.2 %{Delta}k over the 4-month cycle. The core average flux level is {approx}2.4 x 10{sup 15} n/cm{sup 2}s, and test assembly flux level is {approx}2.8 x 10{sup 15} n/cm{sup 2}s. Compared to the metal fuel core, the lower density oxide fuel core requires an average TRU enrichment of 21.8%, which results in a 780 kg TRU loading (as compared to 732 kg for metal) despite a {approx}9% smaller heavy metal inventory. The lower heavy metal inventory increases the burnup reactivity swing by {approx}10% and reduces the flux levels by {approx}8%. Alternative designs were also studied for a LWR-SF TRU feed and a low conversion ratio, including the recycle of the ABTR spent fuel TRU. The lower fissile contents of the LWR-SF TRU relative to the WG-Pu TRU significantly increase the required TRU enrichment of the startup cores to maintain the same cycle length. The even lower fissile fraction of the ABTR spent fuel TRU furt

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

    SciTech Connect (OSTI)

    D. E. Shropshire

    2009-01-01T23:59:59.000Z

    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.

  5. advanced burner reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  6. advanced burner reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

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

    SciTech Connect (OSTI)

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

    2008-02-01T23:59:59.000Z

    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.

  8. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect (OSTI)

    S. T. Khericha

    2007-04-01T23:59:59.000Z

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  9. Supercritical Carbon Dioxide Brayton Cycle Energy Conversion for Sodium-Cooled Fast Reactors/Advanced Burner Reactors

    SciTech Connect (OSTI)

    Sienicki, James J.; Moisseytsev, Anton; Cho, Dae H.; Momozaki, Yoichi; Kilsdonk, Dennis J.; Haglund, Robert C.; Reed, Claude B.; Farmer, Mitchell T. [Argonne National Laboratory 9700 South Cass Avenue, Argonne, Illinois 60439 (United States)

    2007-07-01T23:59:59.000Z

    An optimized supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle power converter has been developed for the 100 MWe (250 MWt) Advanced Burner Test Reactor (ABTR) eliminating the potential for sodium-water reactions and achieving a small power converter and turbine generator building. Cycle and plant efficiencies of 39.1 and 38.3 %, respectively, are calculated for the ABTR core outlet temperature of 510 deg. C. The ABTR S-CO{sub 2} Brayton cycle will incorporate Printed Circuit Heat Exchanger{sup TM} units in the Na-to-CO{sub 2} heat exchangers, high and low temperature recuperators, and cooler. A new sodium test facility is being completed to investigate the potential for transient plugging of narrow sodium channels typical of a Na-to-CO{sub 2} heat exchanger under postulated off-normal or accident conditions. (authors)

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

    SciTech Connect (OSTI)

    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)

    2012-07-01T23:59:59.000Z

    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. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)

  11. Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios.

    SciTech Connect (OSTI)

    Hoffman, E. A.; Yang, W. S.; Hill, R. N.; Nuclear Engineering Division

    2008-05-05T23:59:59.000Z

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond the current fuel database, which is anticipated to be qualified by and for the Advanced Burned Test Reactor. Safety and kinetic parameters were calculated, but a safety analysis was not performed. Development of these designs was required to achieve the primary goal of this study, which was to generate representative fuel cycle mass flows for system studies of ABRs as part of the Global Nuclear Energy Partnership (GNEP). There are slight variations with conversion ratio but the basic ABR configuration consists of 144 fuel assemblies and between 9 and 22 primary control assemblies for both the metal and oxide-fueled cores. Preliminary design studies indicated that it is feasible to design the ABR to accommodate a wide range of conversion ratio by employing different assembly designs and including sufficient control assemblies to accommodate the large reactivity swing at low conversion ratios. The assemblies are designed to fit within the same geometry, but the size and number of fuel pins within each assembly are significantly different in order to achieve the target conversion ratio while still satisfying thermal limits. Current irradiation experience would allow for a conversion ratio of somewhat below 0.75. The fuel qualification for the first ABR should expand this experience to allow for much lower conversion ratios and higher bunrups. The current designs were based on assumptions about the performance of high and very high enrichment fuel, which results in significant uncertainty about the details of the designs. However, the basic fuel cycle performance trends such as conversion ratio and mass flow parameters are less sensitive to these parameters and the current results should provide a good basis for static and dynamic system analysis. The conversion ratio is fundamentally a ratio of the macroscopic cross section of U-238 capture to that of TRU fission. Since the microscopic cross sections only change moderately with fuel design and isotopic concentration for the fast reactor, a specific conversion ratio requires a specific enrichment. The approximate average charge enrichment (TRU/HM) is 14%, 21%, 33%, 56%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the metal-fueled cores. The approximate average charge enrichment is 17%, 25%, 38%, 60%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the oxide-fueled core. For the split batch cores, the maximum enrichment will be somewhat higher. For both the metal and oxide-fueled cores, the reactivity feedback coefficients and kinetics parameters seem reasonable. The maximum single control assembly reactivity faults may be too large for the low conversion ratio designs. The average reactivity of the primary control assemblies was increased, which may cause the maximum reactivity of the central control assembly to be excessive. The values of the reactivity coefficients and kinetics parameters show that some values appear to improve significantly at lower conversion ratios while others appear far less favorable. Detailed safety analysis is required to determine if these designs have adequate safety margins or if appropriate design modifications are required. Detailed system analysis data has been generated for both metal and oxide-fueled core designs over the entire range of potential burner reactors. Additional data has been calculated for a few alternative fuel cycles. The systems data has been summarized in this report and the detailed data will be provided to the systems analysis team so that static and dynamic system analyses can be performed.

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

    SciTech Connect (OSTI)

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

    2011-12-29T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2007-06-30T23:59:59.000Z

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

  14. Advanced Burners and Combustion Controls for Industrial Heat Recovery Systems

    E-Print Network [OSTI]

    Ferri, J. L.

    ADVANCED BURNERS AND COMBUSTION CONTROLS FOR INDUSTRIAL HEAT RECOVERY SYSTEMS J.L.FERRI GTE PRODUCTS CORPORATION TOWANDA, PA ABSTRACT When recuperators are installed on indus trial furnaces, burners and ratio control systems must...ChieVi able not only through design, but also I because the burner internals are all;: ceramic and can wi thstand high tempera~ tures, particularly at low inputs (higih turndown) where the flame front recedes into the burner. A burner test furnace...

  15. advanced burner test: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    burner test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Advanced Burners and Combustion Controls...

  16. Advanced Petrochemical Process Heating with the Pyrocore Burner

    E-Print Network [OSTI]

    Krill, W. V.; Minden, A. C.; Donaldson, L. W. Jr.

    natural gas or refinery process gas and designed to take full advantage of the Pyrocore burner's radiant heat transfer characteristics. This will result in a process heater with design and performance attributes that will be attractive to users...ADVANCED PETROCHEMICAL PROCESS HEATING WITH THE PYROCORE BURNER WAYNE V. KRILL ANDREW C. MINDEN LESLIE W. DONALDSON, JR. Vice President Project Engineer Manager, Process Systems Research Alzeta Corporation Alzeta Corporation Gas Research...

  17. Global Nuclear Energy Partnership Fact Sheet - Develop Advanced...

    Office of Environmental Management (EM)

    Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner Reactors Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner Reactors GNEP will develop...

  18. Advanced Burners and Combustion Controls for Industrial Heat Recovery Systems 

    E-Print Network [OSTI]

    Ferri, J. L.

    1988-01-01T23:59:59.000Z

    When recuperators are installed on industrial furnaces, burners and ratio control systems must continue to operate reliably under a wider range of conditions. Most currently available hot air burners use dilution air to prevent fuel decomposition...

  19. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01T23:59:59.000Z

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  20. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28T23:59:59.000Z

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  1. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01T23:59:59.000Z

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  2. Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy ChinaofSchaefer To: CongestionDevelopment of a downholeReactors | Department of

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

    SciTech Connect (OSTI)

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

    2007-10-01T23:59:59.000Z

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

  4. GNEP Element:Develop Advanced Burner Reactors | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM Flash2011-12 OPAM RevisedFundingEnergy Issues Related toDevelop

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

    SciTech Connect (OSTI)

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

    2010-02-01T23:59:59.000Z

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

  6. PIA - Advanced Test Reactor National Scientific User Facility...

    Energy Savers [EERE]

    Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

  7. Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories

    Office of Legacy Management (LM)

    Radiological Condition of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories Cheswick, Pennsylvania -. -, -- AGENCY: Office of Operational Safety, Department...

  8. DOE - Office of Legacy Management -- Westinghouse Advanced Reactors...

    Office of Legacy Management (LM)

    Advanced Reactors Div Plutonium and Advanced Fuel Labs - PA 10 FUSRAP Considered Sites Site: WESTINGHOUSE ADVANCED REACTORS DIV., PLUTONIUM FUEL LABORATORIES, AND THE ADVANCED FUEL...

  9. Startup burner

    DOE Patents [OSTI]

    Zhao, Jian Lian (Belmont, MA); Northrop, William F. (Ann Arbor, MI); Bosco, Timothy (Dallas, TX); Rizzo, Vincent (Norfolk, MA); Kim, Changsik (Lexington, MA)

    2009-08-18T23:59:59.000Z

    A startup burner for rapidly heating a catalyst in a reformer, as well as related methods and modules, is disclosed.

  10. Advanced Reactor Thermal Hydraulic Modeling | Argonne Leadership...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Advanced Reactor Thermal Hydraulic Modeling PI Name: Paul Fischer PI Email: fischer@mcs.anl.gov Institution: Argonne National Laboratory Allocation Program: INCITE Allocation Hours...

  11. Advanced Reactor Thermal Hydraulic Modeling | Argonne Leadership...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Fischer (ANL), Aleks Obabko (ANL), and Hank Childs (LBNL) Advanced Reactor Thermal Hydraulic Modeling PI Name: Paul Fischer PI Email: fischer@mcs.anl.gov Institution: Argonne...

  12. Advanced Reactor Research and Development Funding Opportunity...

    Broader source: Energy.gov (indexed) [DOE]

    of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and...

  13. Plant maintenance and advanced reactors, 2006

    SciTech Connect (OSTI)

    Agnihotri, Newal (ed.)

    2006-09-15T23:59:59.000Z

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  14. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01T23:59:59.000Z

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  15. Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment

    Broader source: Energy.gov [DOE]

    Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory

  16. STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver

    Broader source: Energy.gov (indexed) [DOE]

    Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low...

  17. Advanced Catalytic Hydrogenation Retrofit Reactor

    SciTech Connect (OSTI)

    Reinaldo M. Machado

    2002-08-15T23:59:59.000Z

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

  18. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15T23:59:59.000Z

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  19. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    BOWEN, W.W.

    1999-11-08T23:59:59.000Z

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  20. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    GANTT, D.A.

    2000-01-12T23:59:59.000Z

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  1. Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01T23:59:59.000Z

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

  2. The advanced test reactor strategic evaluation program

    SciTech Connect (OSTI)

    Buescher, B.J.; Majumdar, D.; Croucher, D.W.

    1989-01-01T23:59:59.000Z

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed.

  3. Uncertainty quantification approaches for advanced reactor analyses.

    SciTech Connect (OSTI)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24T23:59:59.000Z

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  4. Mirror Advanced Reactor Study interim design report

    SciTech Connect (OSTI)

    Not Available

    1983-04-01T23:59:59.000Z

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  5. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

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

    2009-09-01T23:59:59.000Z

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

  6. Beryllium Use in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Glen R. Longhurst

    2007-12-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

  7. Advanced reactor safety research, quarterly report, October-December 1980

    SciTech Connect (OSTI)

    Not Available

    1982-01-01T23:59:59.000Z

    Information is presented concerning advanced reactor core phenomenology; light water reactor severe core damage phenomenology; core debris behavior; containment analysis; elevated temperature design assessment; LMFBR accident delineation; and test and facility technology.

  8. advanced marine reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    fueling; - Te ITB formation - Heat accumulation ? Extension of research area towards reactor-relevant regime Development of ECRF (110 GHz) and N 28 An Advanced Computational...

  9. MELCOR development for existing and advanced reactors

    SciTech Connect (OSTI)

    Summers, R.M.

    1993-12-31T23:59:59.000Z

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions` and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR`s capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed.

  10. ASME Material Challenges for Advanced Reactor Concepts

    SciTech Connect (OSTI)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01T23:59:59.000Z

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2005-10-01T23:59:59.000Z

    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.

  12. Advanced Modularity Design for The MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Advanced Modularity Design for The MIT Pebble Bed Reactor Andrew C. Kadak Department of Nuclear Reactor Technology Institute of Nuclear and New Energy Technology Friendship Hotel, Haidian District Beijing, China September 22-24, 2004 Abstract The future of all reactors will depend on whether they can

  13. LBB application in the US operating and advanced reactors

    SciTech Connect (OSTI)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01T23:59:59.000Z

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  14. Advanced ceramic cladding for water reactor fuel

    SciTech Connect (OSTI)

    Feinroth, H.

    2000-07-01T23:59:59.000Z

    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 {ge}50% would be examined.

  15. Advanced reactor safety research. Quarterly report, July-September 1981

    SciTech Connect (OSTI)

    Not Available

    1982-10-01T23:59:59.000Z

    Sandia National Laboratories, Albuquerque, New Mexico, is conducting the Advanced Reactor Safety Research Program on behalf of the US Nuclear Regulatory Commission (NRC). Sandia has been given the task to investigate seven major areas of interest which are intimately related to over-all NRC needs. These are: core debris behavior - inherent retention; containment analysis; elevated temperature design assessment; LMFBR accident delineation; advanced reactor core phenomenology; light water reactor (LWR) fuel damage phenomenology; and test and facility technology.

  16. Radiation-Induced Segregation and Phase Stability in Candidate Alloys for the Advanced Burner Reactor

    SciTech Connect (OSTI)

    Gary S. Was; Brian D. Wirth

    2011-05-29T23:59:59.000Z

    Major accomplishments of this project were the following: 1) Radiation induced depletion of Cr occurs in alloy D9, in agreement with that observed in austenitic alloys. 2) In F-M alloys, Cr enriches at PAG grain boundaries at low dose (<7 dpa) and at intermediate temperature (400°C) and the magnitude of the enrichment decreases with temperature. 3) Cr enrichment decreases with dose, remaining enriched in alloy T91 up to 10 dpa, but changing to depletion above 3 dpa in HT9 and HCM12A. 4) Cr has a higher diffusivity than Fe by a vacancy mechanism and the corresponding atomic flux of Cr is larger than Fe in the opposite direction to the vacancy flux. 5) Cr concentration at grain boundaries decreases as a result of vacancy transport during electron or proton irradiation, consistent with Inverse Kirkendall models. 6) Inclusion of other point defect sinks into the KLMC simulation of vacancy-mediated diffusion only influences the results in the low temperature, recombination dominated regime, but does not change the conclusion that Cr depletes as a result of vacancy transport to the sink. 7) Cr segregation behavior is independent of Frenkel pair versus cascade production, as simulated for electron versus proton irradiation conditions, for the temperatures investigated. 8) The amount of Cr depletion at a simulated planar boundary with vacancy-mediated diffusion reaches an apparent saturation value by about 1 dpa, with the precise saturation concentration dependent on the ratio of Cr to Fe diffusivity. 9) Cr diffuses faster than Fe by an interstitial transport mechanism, and the corresponding atomic flux of Cr is much larger than Fe in the same direction as the interstitial flux. 10) Observed experimental and computational results show that the radiation induced segregation behavior of Cr is consistent with an Inverse Kirkendall mechanism.

  17. Front Burner- Issue 14

    Broader source: Energy.gov [DOE]

    The Cybersecurity Front Burner Issue No. 14 addresses the 2013 National Cybersecurity Awareness Month (NCSAM) Campaign and Phishing Scams.

  18. Front Burner- Issue 15

    Broader source: Energy.gov [DOE]

    The Cybersecurity Front Burner Issue No. 15 addresses the DOE eSCRM Program and Secure Online Shopping.

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

    SciTech Connect (OSTI)

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

    2014-04-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01T23:59:59.000Z

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

  1. Reactor assessments of advanced bumpy torus configurations

    SciTech Connect (OSTI)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1983-01-01T23:59:59.000Z

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  2. advanced reactor design: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  3. advanced reactor designs: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  4. advanced reactor systems: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  5. advanced reactors advanced: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  6. Front Burner- Issue 13

    Broader source: Energy.gov [DOE]

    The Cybersecurity Front Burner Issue No. 13 contained a message from the Associate Chief Information Officer (ACIO) for Cybersecurity as well as a listing of recommended cybersecurity practices.

  7. Front Burner- Issue 18

    Broader source: Energy.gov [DOE]

    The Cybersecurity Front Burner Issue No. 18 addresses keeping kids safe on the Internet, cyber crime, and DOE Cyber awareness and training initiatives.

  8. Front Burner- Issue 16

    Broader source: Energy.gov [DOE]

    The Cybersecurity Front Burner Issue No. 16 addresses Malware, the Worst Passwords of 2013, and the Flat Stanley and Stop.Think.Connect. Campaign.

  9. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect (OSTI)

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01T23:59:59.000Z

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  10. Combustor burner vanelets

    DOE Patents [OSTI]

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

    2012-02-14T23:59:59.000Z

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

  11. E-Print Network 3.0 - advanced research reactor Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

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

    SciTech Connect (OSTI)

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

    2012-05-31T23:59:59.000Z

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded

  13. Advanced Test Reactor Capabilities and Future Irradiation Plans

    SciTech Connect (OSTI)

    Frances M. Marshall

    2006-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.

  14. Advanced Nuclear Reactors | Department of Energy

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041cloth DocumentationProducts (VAP) VAP7-0973 1BP-14 PowerAdvanced Modeling &Advanced Nuclear

  15. advanced test reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    test reactor First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Advancing Toward Test Automation through...

  16. advanced-gas-cooled-nuclear-reactor materials evaluation: Topics...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    advanced-gas-cooled-nuclear-reactor materials evaluation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index...

  17. Examination of the legal mechanisms to regulate advanced fision reactors

    SciTech Connect (OSTI)

    Brinig, M.F.; Repici, D.J.

    1988-12-01T23:59:59.000Z

    The George Mason University School of Law (GMUSL) located in Northern Virginia, and its subcontractor, The John Francis Company, Inc., of Fairfax, Virginia, conducted a study for the Department of Energy's Office of Nuclear Energy which examined the legal mechanisms for the regulation of advanced fision reactors. This report presents the research and findings conducted under that study.

  18. Advanced Reactor Technology Documents | Department of Energy

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed off Energy.gov. Are you0 ARRA Newsletters 20103-03 AUDITProductsletterInitiatives »Nuclear Reactor

  19. Advanced reactor safety research. Quarterly report, October-December 1981. Volume 20

    SciTech Connect (OSTI)

    Not Available

    1983-08-01T23:59:59.000Z

    Information is presented concerning the inherent retention of core debris following a severe reactor accident; containment analysis for LWR and LMFBR type reactors; LMFBR accident delineation; advanced reactor core phenomenology; LWR damaged fuel phenomenology; and ACRR status.

  20. Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award...

    Office of Environmental Management (EM)

    Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award October 16, 2014 - 5:24pm Addthis Developed...

  1. The impact of passive safety systems on desirability of advanced light water reactors

    E-Print Network [OSTI]

    Eul, Ryan C

    2006-01-01T23:59:59.000Z

    This work investigates whether the advanced light water reactor designs with passive safety systems are more desirable than advanced reactor designs with active safety systems from the point of view of uncertainty in the ...

  2. Pulverized coal burner

    DOE Patents [OSTI]

    Sivy, Jennifer L. (Alliance, OH); Rodgers, Larry W. (Canton, OH); Koslosy, John V. (Akron, OH); LaRue, Albert D. (Uniontown, OH); Kaufman, Keith C. (Canton, OH); Sarv, Hamid (Canton, OH)

    1998-01-01T23:59:59.000Z

    A burner having lower emissions and lower unburned fuel losses by implementing a transition zone in a low NO.sub.x burner. The improved burner includes a pulverized fuel transport nozzle surrounded by the transition zone which shields the central oxygen-lean fuel devolatilization zone from the swirling secondary combustion air. The transition zone acts as a buffer between the primary and the secondary air streams to improve the control of near-burner mixing and flame stability by providing limited recirculation regions between primary and secondary air streams. These limited recirculation regions transport evolved NO.sub.x back towards the oxygen-lean fuel pyrolysis zone for reduction to molecular nitrogen. Alternate embodiments include natural gas and fuel oil firing.

  3. Pulverized coal burner

    DOE Patents [OSTI]

    Sivy, J.L.; Rodgers, L.W.; Koslosy, J.V.; LaRue, A.D.; Kaufman, K.C.; Sarv, H.

    1998-11-03T23:59:59.000Z

    A burner is described having lower emissions and lower unburned fuel losses by implementing a transition zone in a low NO{sub x} burner. The improved burner includes a pulverized fuel transport nozzle surrounded by the transition zone which shields the central oxygen-lean fuel devolatilization zone from the swirling secondary combustion air. The transition zone acts as a buffer between the primary and the secondary air streams to improve the control of near-burner mixing and flame stability by providing limited recirculation regions between primary and secondary air streams. These limited recirculation regions transport evolved NO{sub x} back towards the oxygen-lean fuel pyrolysis zone for reduction to molecular nitrogen. Alternate embodiments include natural gas and fuel oil firing. 8 figs.

  4. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2004-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  5. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    Grover, S.B.

    2004-10-06T23:59:59.000Z

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

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

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2007-09-01T23:59:59.000Z

    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.

  7. Burner control system

    SciTech Connect (OSTI)

    Cade, P.J.

    1981-01-06T23:59:59.000Z

    A burner control apparatus for use with a furnace installation that has an operating control to produce a request for burner operation, a flame sensor to produce a signal when flame is present in the monitored combustion chamber, and one or more devices for control of ignition and/or fuel flow. The burner control apparatus comprises lockout apparatus for de-energizing the control apparatus, a control device for actuating the ignition and/or fuel control devices, and a timing circuit that provides four successive and partially overlapping timing intervals of precise relation, including a purge timing interval, a pilot ignition interval, and a main fuel ignition interval. The present invention further includes a burner control system which verifies the proper operation of certain sensors in a burner or furnace including particularly the air flow sensor. Additionally, the present system also prevents an attempt to ignite a burner if a condition is detected which indicates that the air flow sensor has been bypassed or wedged in the actuated position.

  8. Low NO.sub.x burner system

    DOE Patents [OSTI]

    Kitto, Jr., John B. (North Canton, OH); Kleisley, Roger J. (Plain Twp., Stark County, OH); LaRue, Albert D. (Summit, OH); Latham, Chris E. (Knox Twp., Columbiana County, OH); Laursen, Thomas A. (Canton, OH)

    1993-01-01T23:59:59.000Z

    A low NO.sub.x burner system for a furnace having spaced apart front and rear walls, comprises a double row of cell burners on each of the front and rear walls. Each cell burner is either of the inverted type with a secondary air nozzle spaced vertically below a coal nozzle, or the non-inverted type where the coal nozzle is below the secondary air port. The inverted and non-inverted cells alternate or are provided in other specified patterns at least in the lower row of cells. A small percentage of the total air can be also provided through the hopper or hopper throat forming the bottom of the furnace, or through the boiler hopper side walls. A shallow angle impeller design also advances the purpose of the invention which is to reduce CO and H.sub.2 S admissions while maintaining low NO.sub.x generation.

  9. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15T23:59:59.000Z

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  10. Advanced Reactor Technologies | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuels DataEnergyDepartment ofATVM LoanActiveMission »Advanced ModelingNuclear

  11. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect (OSTI)

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01T23:59:59.000Z

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  12. Abnormal operating procedures for ATR (Advanced Test Reactor's) experiment loops

    SciTech Connect (OSTI)

    Auflick, J.L.

    1989-09-01T23:59:59.000Z

    This paper outlines the background from the TMI accident which resulted in the definition and development of function-oriented procedures. It also explains how function-oriented procedures were applied in a task for the Advanced Test Reactor's (ATR) NR experiment loops. Human performance design discrepancies were identified for existing procedures, and were corrected by upgrading them according to current NRC and DOE standards. Finally, specific recommendations are made with respect to future ATR control room and loop improvements, as they relate to the revision of operating procedures within INEL's power reactor program. 8 refs., 4 figs.

  13. Integrated intelligent systems in advanced reactor control rooms

    SciTech Connect (OSTI)

    Beckmeyer, R.R.

    1989-01-01T23:59:59.000Z

    An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

  14. Advanced High Temperature Reactor Neutronic Core Design

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

  15. DOE/NE robotics for advanced reactors

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    This document details activities during this reporting period. The Michigan group has developed, built, and tested a general purpose interface circuit for DC motors and encoders. This interface is based on an advanced microchip, the HCTL 1100 manufactured by Hewlett Packard. The HCTL 1100 can be programmed by a host computer in real-time, allowing sophisticated motion control for DC motors. At the University of Florida, work on modeling the details of the seismic isolators and the jack mechanism has been completed. A separate 3D solid view of the seismic isolator floor, with the full set of isolators shown in detail, has been constructed within IGRIP. ORNL led the robotics team at the ALMR review meeting. Discussions were held with General Electric (GE) engineers and contractors on the robotic needs for the ALMR program. The Tennessee group has completed geometric modeling of the Andros Mark VI mobile platform with two fixed tracks and for articulated tracks, the give degree-of-freedom manipulator and its end-effector, and two cameras. A graphical control of panel was developed which allow the user to operate the simulated robot. The University of Texas team visited ORNL to complete the implementation of computed-torque controller on the CESARm manipulator. This controller was previously developed and computer simulations were carried out specifically for the CESARm robot.

  16. Advances in process intensification through multifunctional reactor engineering

    SciTech Connect (OSTI)

    O'Hern, T. J.

    2012-03-01T23:59:59.000Z

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes critical to process intensification and implementation in commercial applications. Physics of the heat and mass transfer and chemical kinetics and how these processes are ultimately scaled were investigated. Specifically, we progressed the knowledge and tools required to scale a multifunctional reactor for acid-catalyzed C4 paraffin/olefin alkylation to industrial dimensions. Understanding such process intensification strategies is crucial to improving the energy efficiency and profitability of multifunctional reactors, resulting in a projected energy savings of 100 trillion BTU/yr by 2020 and a substantial reduction in the accompanying emissions.

  17. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2012-10-01T23:59:59.000Z

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  18. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  19. Development of advanced strain diagnostic techniques for reactor environments.

    SciTech Connect (OSTI)

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01T23:59:59.000Z

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  20. Ultralean low swirl burner

    DOE Patents [OSTI]

    Cheng, Robert K. (Kensington, CA)

    1998-01-01T23:59:59.000Z

    A novel burner and burner method has been invented which burns an ultra lean premixed fuel-air mixture with a stable flame. The inventive burning method results in efficient burning and much lower emissions of pollutants such as oxides of nitrogen than previous burners and burning methods. The inventive method imparts weak swirl (swirl numbers of between about 0.01 to 3.0) on a fuel-air flow stream. The swirl, too small to cause recirculation, causes an annulus region immediately inside the perimeter of the fuel-air flow to rotate in a plane normal to the axial flow. The rotation in turn causes the diameter of the fuel-air flow to increase with concomitant decrease in axial flow velocity. The flame stabilizes where the fuel-air mixture velocity equals the rate of burning resulting in a stable, turbulent flame.

  1. Ultralean low swirl burner

    DOE Patents [OSTI]

    Cheng, R.K.

    1998-04-07T23:59:59.000Z

    A novel burner and burner method has been invented which burns an ultra lean premixed fuel-air mixture with a stable flame. The inventive burning method results in efficient burning and much lower emissions of pollutants such as oxides of nitrogen than previous burners and burning methods. The inventive method imparts weak swirl (swirl numbers of between about 0.01 to 3.0) on a fuel-air flow stream. The swirl, too small to cause recirculation, causes an annulus region immediately inside the perimeter of the fuel-air flow to rotate in a plane normal to the axial flow. The rotation in turn causes the diameter of the fuel-air flow to increase with concomitant decrease in axial flow velocity. The flame stabilizes where the fuel-air mixture velocity equals the rate of burning resulting in a stable, turbulent flame. 11 figs.

  2. Freese-casting as a Novel Manufacturing Process for Fast Reactor Fuels

    SciTech Connect (OSTI)

    Wegst, Ulrike G.K.; Allen, Todd; Sridharan, Kumar

    2014-04-07T23:59:59.000Z

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

  3. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect (OSTI)

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

    2012-03-01T23:59:59.000Z

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    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.

  5. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13T23:59:59.000Z

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  6. Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants

    E-Print Network [OSTI]

    Anitescu, Mihai

    Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 by G. Palmiotti, J. Cahalan, P. Pfeiffer, T;2 ANL-AFCI-168 Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants G

  7. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect (OSTI)

    Timothy O’Hern, Lindsey Evans, Jim Miller, Marcia Cooper, John Torczynski, Donovan Pena, and Walt Gill, SNL

    2011-02-01T23:59:59.000Z

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in other technologies.

  8. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect (OSTI)

    Timothy O’Hern, Lindsey Evans, Jim Miller, Marcia Cooper, John Torczynski, Donovan Pena, and Walt Gill, SNL, Will Groten, Arvids Judzis, Richard Foley, Larry Smith, and Will Cross, CR& L / CDTECH; T. Vogt, Lummus Technology / CDTECH.

    2011-06-27T23:59:59.000Z

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in other technologies.

  9. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18T23:59:59.000Z

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

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

    SciTech Connect (OSTI)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

    2011-06-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

    2012-08-01T23:59:59.000Z

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

  13. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect (OSTI)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11T23:59:59.000Z

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

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

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2008-10-01T23:59:59.000Z

    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.

  15. Replacement of the Advanced Test Reactor control room

    SciTech Connect (OSTI)

    Durney, J.L.; Klingler, W.B. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-01-01T23:59:59.000Z

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor. 1 fig., 1 tab.

  16. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect (OSTI)

    Michael Swanson; Daniel Laudal

    2008-03-31T23:59:59.000Z

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

  17. A Novel Approach to Material Development for Advanced Reactor Systems

    SciTech Connect (OSTI)

    Was, G.S.; Atzmon, M.; Wang, L.

    2000-06-27T23:59:59.000Z

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use of low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steel without the effects of RIS. Third, to initiate irradiation of reactor pressure vessel steel and Zircaloy. In year 1 quarter 3, the project goal was to complete irradiation of model alloys of RPV steels for a range of doses and begin sample characterization. We also planned to prepare samples for microstructure isolation in stainless steels, and to identify sources of Zircaloy for irradiation and characterization.

  18. Advanced reactor safety research quarterly report, January-March 1982. Vol. 21

    SciTech Connect (OSTI)

    Not Available

    1983-08-01T23:59:59.000Z

    Information is presented concerning core debris behavior (inherent retention); containment analysis; elevated temperature design assessment; Clinch River risk assessment study; advanced reactor core phenomenology; LWR damaged fuel relocation phenomenology; and Annular Core Research Reactor facilities and operation.

  19. Evolutionary/advanced light water reactor data report

    SciTech Connect (OSTI)

    NONE

    1996-02-09T23:59:59.000Z

    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.

  20. Metal fires and their implications for advanced reactors.

    SciTech Connect (OSTI)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01T23:59:59.000Z

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.

  1. Distillation and Dehydro Reactors Advanced Process Conrol Freeport Texas PLant

    E-Print Network [OSTI]

    Eisele, D.

    2014-01-01T23:59:59.000Z

    Distillation and Dehydro Reactors Advanced Process Control Freeport Texas Plant ESL-IE-14-05-16 Proceedings of the Thrity-Sixth Industrial Energy Technology Conference New Orleans, LA. May 20-23, 2014 G-KTI, Polyamide and Intermediates Distillation...-Sixth Industrial Energy Technology Conference New Orleans, LA. May 20-23, 2014 G-KTI, Polyamide and Intermediates Distillation APC – Control Matrix 6/2/2014 INTERNAL; CONFIDENTIAL 3 ESL-IE-14-05-16 Proceedings of the Thrity-Sixth Industrial Energy Technology...

  2. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect (OSTI)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20T23:59:59.000Z

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.

  3. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect (OSTI)

    Frances M. Marshall

    2005-04-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

  4. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    selected as part of the Generation IV reactors .. - 4 -The development of Generation IV fast reactors can make aconcepts selected for the Generation IV reactors, three,

  5. Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators

    SciTech Connect (OSTI)

    Jeffrey C. Joe; Johanna H. Oxstrand

    2014-06-01T23:59:59.000Z

    The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks. We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.

  6. DEVELOPMENT OF A NOVEL RADIATIVELY/CONDUCTIVELY STABILIZED BURNER FOR SIGNIFICANT REDUCTION OF NOx EMISSIONS AND FOR ADVANCING THE MODELING AND UNDERSTANDING OF PULVERIZED COAL COMBUSTION AND EMISSIONS

    SciTech Connect (OSTI)

    Noam Lior; Stuart W. Churchill

    2003-10-01T23:59:59.000Z

    The primary objective of the proposed study was the study and analysis of, and design recommendations for, a novel radiatively-conductively stabilized combustion (RCSC) process for pulverized coal, which, based on our prior studies with both fluid fuels and pulverized coal, holds a high promise to reduce NO{sub x} production significantly. We have primarily engaged in continuing and improving our process modeling and analysis, obtained a large amount of quantitative information about the effects of the major parameters on NO{sub x} production, conducted an extensive exergy analysis of the process, evaluated the practicalities of employing the Radiatively-Conductively Stabilized Combustor (RCSC) to large power and heat plants, and improved the experimental facility. Prior experimental work has proven the feasibility of the combustor, but slagging during coal combustion was observed and should be dealt with. The primary outcomes and conclusions from the study are: (1) we developed a model and computer program that represents the pulverized coal combustion in the RCSC, (2) the model predicts that NO{sub x} emissions can be reduced by a number of methods, detailed in the report. (3) the exergy analysis points out at least a couple of possible ways to improve the exergetic efficiency in this combustor: increasing the effectiveness of thermal feedback, and adjusting the combustor mixture exit location, (4) because of the low coal flow rates necessitated in this study to obtain complete combustion in the burner, the size of a burner operating under the considered conditions would have to be up to an order of magnitude, larger than comparable commercial burners, but different flow configurations of the RCSC can yield higher feed rates and smaller dimensions, and should be investigated. Related to this contract, eleven papers were published in journals and conference proceedings, and ten invited presentations were given at university and research institutions, as well as at the Gordon Conference on Modern Development in Thermodynamics. The results obtained are very encouraging for the development of the RCSC as a commercial burner for significant reduction of NO{sub x} emissions, and highly warrants further study and development.

  7. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover; John Maki; David Petti

    2010-10-01T23:59:59.000Z

    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). 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 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 completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during irradiation will be discussed and the results of the experiment will be presented. The second experiment (AGR-2) is currently being assembled, and the status as well as the new fuel and irradiation conditions for that experiment will also be discussed.

  8. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Gas Expansion Module Gas-cooled Fast Reactor High Enrichedfast reactors: gas-cooled fast reactor (GFR), sodium-cooledderived from the Gas cooled Fast Reactor (GFR). This core

  9. Advanced Neutron Source reactor control and plant protection systems design

    SciTech Connect (OSTI)

    Anderson, J.L.; Battle, R.E.; March-Leuba, J. (Oak Ridge National Lab., TN (United States)); Khayat, M.I. (Tennessee Univ., Knoxville, TN (United States))

    1992-01-01T23:59:59.000Z

    This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges.

  10. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01T23:59:59.000Z

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  11. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect (OSTI)

    Michael L. Swanson

    2005-08-30T23:59:59.000Z

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was 50 hours of gasification on a petroleum coke from the Hunt Oil Refinery and an additional 73 hours of operation on a high-ash coal from India. Data from these tests indicate that while acceptable fuel gas heating value was achieved with these fuels, the transport gasifier performs better on the lower-rank feedstocks because of their higher char reactivity. Comparable carbon conversions have been achieved at similar oxygen/coal ratios for both air-blown and oxygen-blown operation for each fuel; however, carbon conversion was lower for the less reactive feedstocks. While separation of fines from the feed coals is not needed with this technology, some testing has suggested that feedstocks with higher levels of fines have resulted in reduced carbon conversion, presumably due to the inability of the finer carbon particles to be captured by the cyclones. These data show that these low-rank feedstocks provided similar fuel gas heating values; however, even among the high-reactivity low-rank coals, the carbon conversion did appear to be lower for the fuels (brown coal in particular) that contained a significant amount of fines. The fuel gas under oxygen-blown operation has been higher in hydrogen and carbon dioxide concentration since the higher steam injection rate promotes the water-gas shift reaction to produce more CO{sub 2} and H{sub 2} at the expense of the CO and water vapor. However, the high water and CO{sub 2} partial pressures have also significantly reduced the reaction of (Abstract truncated)

  12. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect (OSTI)

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  13. Advanced Reactor Safety Research Division quarterly progress report, January 1-March 31, 1981

    SciTech Connect (OSTI)

    Cerbone, R.J.; Ginsberg, T.; Guppy, J.G.; Sastre, C.

    1981-05-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  14. Advanced Reactor Safety Research Division. Quarterly progress report, July 1-September 30, 1980

    SciTech Connect (OSTI)

    Ramano, A.J. (comp.)

    1980-11-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  15. Advanced Reactor Safety Research Division quarterly progress report, 1 October-31 December 1980

    SciTech Connect (OSTI)

    Cerbone, R.J.; Ginsberg, T.; Guppy, J.G.; Sastre, C.

    1981-02-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, LMFBR Safety Experiments, SSC Code Development, and Fast Reactor Safety Code Validation.

  16. Advanced Reactor Safety Research Division. Quarterly progress report, July 1-September 30, 1979

    SciTech Connect (OSTI)

    Romano, A.J.

    1980-01-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  17. Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect (OSTI)

    Romano, A.J.

    1980-01-01T23:59:59.000Z

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  18. Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    ARIES stands for Advanced Reactor Innovation Evaluation Study. It is a program and a team that explores the commercial potential of fusion as an energy resource. Though it is a multi-institutional program, ARIES is led by the University of California at San Diego. ARIES studies both magnetic fusion energy (MFE) and inertial fusion energy (IFE), using an approach that integrates theory, experiments, and technology. The ARIES team proposes fusion reactor designs and works to understand how technology, materials and plasma physics processes interact and influence each other. A 2005 report to the Fusion Energy Sciences Advisory Committee ("Scientific Challenges, Opportunities, and Priorities for the U.S. Fusion Energy Sciences Program") noted on page 98 an example of the importance of this materials properties aspect: "For instance, effects on plasma edge by various plasma facing materials and effects on various plasma stabilization and control techniques by highly conducting liquid metal blankets are being considered by physicists." This web page is an archive of material properties collected here for the use of the ARIES Fusion Power Plant Studies Team.

  19. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Fuels for sodium-cooled fast reactors: US perspective.Pitch to Diameter Sodium-cooled Fast Reactor Simple Movingreactor (GFR), sodium-cooled fast reactor (SFR) and lead-

  20. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Reports. 1999. IAEA, Fast Reactor Database 2006 Update, inof the Breed/Burn Fast Reactor Concept, in Department ofof Ultra-long Life Fast Reactor Core Concept, in Physor

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

    SciTech Connect (OSTI)

    Sharp, G.L.; McCracken, R.T.

    2003-05-13T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Gregg L. Sharp; R. T. McCracken

    2003-06-01T23:59:59.000Z

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

  3. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect (OSTI)

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

    2012-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives cannot be met using the INL facilities. The ATR NSUF program includes a robust education program enabling students to participate in their research at INL and the partner facilities, attend the ATR NSUF annual User Week, and compete for prizes at sponsored conferences. Development of additional research capabilities is also a key component of the ATR NSUF Program; researchers are encouraged to propose research projects leading to these enhanced capabilities. Some ATR irradiation experiment projects irradiate more specimens than are tested, resulting in irradiated materials available for post irradiation examination by other researchers. These “extra” specimens comprise the ATR NSUF Sample Library. This presentation will highlight the ATR NSUF Sample Library and the process open to researchers who want to access these materials and how to propose research projects using them. This presentation will provide the current status of all the ATR NSUF Program elements. Many of these were not envisioned in 2007, when DOE established the ATR NSUF.

  4. advanced passive reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Thomas 2006-01-01 12 Radiation Hardness of Passive Fibre Optic Components for the Future Thermonuclear Fusion Reactor CiteSeer Summary: thermon uclearfusion reactor ITER will...

  5. advanced fusion reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Collaborators 7 China To Build Its Own Fusion Reactor ENERGY TECH Plasma Physics and Fusion Websites Summary: Thermonuclear Experimental Reactor project reached agreement in...

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-09-01T23:59:59.000Z

    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 addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

  7. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1992-04-01T23:59:59.000Z

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  8. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect (OSTI)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01T23:59:59.000Z

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

  9. Radial lean direct injection burner

    DOE Patents [OSTI]

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

    2012-09-04T23:59:59.000Z

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

  10. Catalyzed Ceramic Burner Material

    SciTech Connect (OSTI)

    Barnes, Amy S., Dr.

    2012-06-29T23:59:59.000Z

    Catalyzed combustion offers the advantages of increased fuel efficiency, decreased emissions (both NOx and CO), and an expanded operating range. These performance improvements are related to the ability of the catalyst to stabilize a flame at or within the burner media and to combust fuel at much lower temperatures. This technology has a diverse set of applications in industrial and commercial heating, including boilers for the paper, food and chemical industries. However, wide spread adoption of catalyzed combustion has been limited by the high cost of precious metals needed for the catalyst materials. The primary objective of this project was the development of an innovative catalyzed burner media for commercial and small industrial boiler applications that drastically reduce the unit cost of the catalyzed media without sacrificing the benefits associated with catalyzed combustion. The scope of this program was to identify both the optimum substrate material as well as the best performing catalyst construction to meet or exceed industry standards for durability, cost, energy efficiency, and emissions. It was anticipated that commercial implementation of this technology would result in significant energy savings and reduced emissions. Based on demonstrated achievements, there is a potential to reduce NOx emissions by 40,000 TPY and natural gas consumption by 8.9 TBtu in industries that heavily utilize natural gas for process heating. These industries include food manufacturing, polymer processing, and pulp and paper manufacturing. Initial evaluation of commercial solutions and upcoming EPA regulations suggests that small to midsized boilers in industrial and commercial markets could possibly see the greatest benefit from this technology. While out of scope for the current program, an extension of this technology could also be applied to catalytic oxidation for volatile organic compounds (VOCs). Considerable progress has been made over the course of the grant period in accomplishing these objectives. Our work in the area of Pd-based, methane oxidation catalysts has led to the development of highly active catalysts with relatively low loadings of Pd metal using proprietary coating methods. The thermal stability of these Pd-based catalysts were characterized using SEM and BET analyses, further demonstrating that certain catalyst supports offer enhanced stability toward both PdO decomposition and/or thermal sintering/growth of Pd particles. When applied to commercially available fiber mesh substrates (both metallic and ceramic) and tested in an open-air burner, these catalyst-support chemistries showed modest improvements in the NOx emissions and radiant output compared to uncatalyzed substrates. More significant, though, was the performance of the catalyst-support chemistries on novel media substrates. These substrates were developed to overcome the limitations that are present with commercially available substrate designs and increase the gas-catalyst contact time. When catalyzed, these substrates demonstrated a 65-75% reduction in NOx emissions across the firing range when tested in an open air burner. In testing in a residential boiler, this translated into NOx emissions of <15 ppm over the 15-150 kBtu/hr firing range.

  11. Acid gas burner

    SciTech Connect (OSTI)

    Polak, B.

    1991-04-23T23:59:59.000Z

    This patent describes a burner for combusting a waste gas. It comprises a throat section; a fire tube downstream from the throat section in communication therewith; an air duct section upstream from the throat section in communication therewith; a centrally located nozzle means for introduction of a fuel in the throat section in a downstream direction toward the fire tube; means upstream from the throat section for forming a downstream directed swirling combustion air stream substantially in an annular ring along the sidewalls of the throat section; and means for introducing a waste gas stream into the throat section downstream of the nozzle means in a forwardly biased but swirling direction opposite to that of the swirling combustion air stream.

  12. advanced fission reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Daya Bay reactor core, and compare its with those calculated by the commercial reactor simulation program SCIENCE, which is used by the Daya Bay nuclear power plant, and the...

  13. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    SciTech Connect (OSTI)

    Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

    1992-01-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs.

  14. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    SciTech Connect (OSTI)

    T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

    2009-05-01T23:59:59.000Z

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.

  15. Licensing and Deployment of Advanced Reactors Andrew C. Kadak, Ph.D.

    E-Print Network [OSTI]

    of advanced reactors of the type proposed for Generation IV will require a new strategy for licensing since many of the proposed Generation IV technologies include concepts such as high temperature gas

  16. Advancing Small Modular Reactors: How We're Supporting Next-Gen...

    Energy Savers [EERE]

    Nuclear Energy Technology Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology December 12, 2013 - 4:00pm Addthis The basics of small modular...

  17. advanced nuclear reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  18. advanced pressurized reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  19. advanced thermal reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  20. advanced reactor analyses: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  1. advanced fast reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  2. advanced water reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  3. advanced integral reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  4. advanced reactor study: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  5. advanced reactors coupled: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  6. advanced reactor instrumentation: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  7. advanced reactor mixed: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  8. advanced reactors transition: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  9. advanced hanaro reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  10. advanced candu reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  11. advanced power reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  12. advanced reactor development: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  13. advanced nuclear reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  14. advanced thermal reactor fugen: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  15. Criterion for burner design in thermal weed control 

    E-Print Network [OSTI]

    Gonzalez, Telca Marisa

    2001-01-01T23:59:59.000Z

    A covered infrared burner was designed and constructed so that it could be compared to an open-flame burner. Two covered burners, a high configuration and a low configuration, were constructed. A low configuration covered infrared burner, high...

  16. DEVELOPMENT OF A LOW PRESSURE, AIR ATOMIZED OIL BURNER WITH HIGH ATOMIZER AIR FLOW

    SciTech Connect (OSTI)

    BUTCHER,T.A.

    1998-01-01T23:59:59.000Z

    This report describes technical advances made to the concept of a low pressure, air atomized oil burner for home heating applications. Currently all oil burners on the market are of the pressure atomized, retention head type. These burners have a lower firing rate limit of about 0.5 gallons per hour of oil, due to reliability problems related to small flow passage sizes. High pressure air atomized burners have been shown to be one route to avoid this problem but air compressor cost and reliability have practically eliminated this approach. With the low pressure air atomized burner the air required for atomization can be provided by a fan at 5--8 inches of water pressure. A burner using this concept, termed the Fan-Atomized Burner or FAB has been developed and is currently being commercialized. In the head of the FAB, the combustion air is divided into three parts, much like a conventional retention head burner. This report describes development work on a new concept in which 100% of the air from the fan goes through the atomizer. The primary advantage of this approach is a great simplification of the head design. A nozzle specifically sized for this concept was built and is described in the report. Basic flow pressure tests, cold air velocity profiles, and atomization performance have been measured. A burner head/flame tube has been developed which promotes a torroidal recirculation zone near the nozzle for flame stability. The burner head has been tested in several furnace and boiler applications over the tiring rate range 0.2 to 0.28 gallons per hour. In all cases the burner can operate with very low excess air levels (under 10%) without producing smoke. Flue gas NO{sub x} concentration varied from 42 to 62 ppm at 3% 0{sub 2}. The concept is seen as having significant potential and planned development efforts are discussed.

  17. advanced reactors division: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    fusion reactor: another way round to tackle the problem is the reduction of the neutron flux and subsequent material ... Zucchetti, Massimo 11 Recovery of Carbon Dioxide in...

  18. advanced converter reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    CERN Preprints Summary: The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium tritium burning...

  19. advanced reactor licensing: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics...

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

    SciTech Connect (OSTI)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11T23:59:59.000Z

    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.

  1. Burner ignition system

    DOE Patents [OSTI]

    Carignan, Forest J. (Bedford, MA)

    1986-01-21T23:59:59.000Z

    An electronic ignition system for a gas burner is battery operated. The battery voltage is applied through a DC-DC chopper to a step-up transformer to charge a capacitor which provides the ignition spark. The step-up transformer has a significant leakage reactance in order to limit current flow from the battery during initial charging of the capacitor. A tank circuit at the input of the transformer returns magnetizing current resulting from the leakage reactance to the primary in succeeding cycles. An SCR in the output circuit is gated through a voltage divider which senses current flow through a flame. Once the flame is sensed, further sparks are precluded. The same flame sensor enables a thermopile driven main valve actuating circuit. A safety valve in series with the main gas valve responds to a control pressure thermostatically applied through a diaphragm. The valve closes after a predetermined delay determined by a time delay orifice if the pilot gas is not ignited.

  2. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect (OSTI)

    NONE

    1993-09-15T23:59:59.000Z

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  3. DECAY HEAT CONDITIONS OF CURRENT AND NEXT GENERATION REACTORS

    E-Print Network [OSTI]

    Choe, JongSoo 1985-

    2012-05-04T23:59:59.000Z

    expects to operate the VHTR by 2021 (NGNP, A Report to Congress, 2008). Advanced Burner 3 Reactors (ABR) are Sodium-Cooled Fast Reactors (SFR) which are fast neutron spectrum and closed fuel cycle system reactors. Its management of actinides... enriched uranium dioxide(UO2 ) less than 5 wt% and gadolinia-uranium dioxide(Gd,UO2). The cladding material is ZIRLO which is a zirconium based alloy for improved corrosion resistance (US-APWR, 2011). The operation power of the ABWR is 3926 MWth. Its...

  4. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect (OSTI)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01T23:59:59.000Z

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  5. RENEWABLES RESEARCH Boiler Burner Energy System Technology

    E-Print Network [OSTI]

    RENEWABLES RESEARCH Boiler Burner Energy System Technology (BBEST) for Firetube Boilers PIER Renewables Research September 2010 The Issue Researchers at Altex Technologies Corporation in Sunnyvale, industrial combined heat and power (CHP) boiler burner energy system technology ("BBEST"). Their research

  6. advanced light reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Light-to-surface distances overwhelm Gordon, Scott 11 Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor MIT - DSpace Summary: A design for a large,...

  7. Advanced fuel fusion reactors: towards a zero-waste option

    E-Print Network [OSTI]

    Zucchetti, Massimo

    Low activation materials are only a partial response to the requirement of a really environmentally sound fusion reactor: another way round to tackle the problem is the reduction of the neutron flux and subsequent material ...

  8. Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

    SciTech Connect (OSTI)

    Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1994-06-01T23:59:59.000Z

    Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

  9. Enhanced Combustion Low NOx Pulverized Coal Burner

    SciTech Connect (OSTI)

    Ray Chamberland; Aku Raino; David Towle

    2006-09-30T23:59:59.000Z

    For more than two decades, ALSTOM Power Inc. (ALSTOM) has developed a range of low cost, in-furnace technologies for NOx emissions control for the domestic U.S. pulverized coal fired boiler market. This includes ALSTOM's internally developed TFS 2000 firing system, and various enhancements to it developed in concert with the U.S. Department of Energy (DOE). As of 2004, more than 200 units representing approximately 75,000 MWe of domestic coal fired capacity have been retrofit with ALSTOM low NOx technology. Best of class emissions range from 0.18 lb/MMBtu for bituminous coals to 0.10 lb/MMBtu for subbituminous coals, with typical levels at 0.24 lb/MMBtu and 0.13 lb/MMBtu, respectively. Despite these gains, NOx emissions limits in the U.S. continue to ratchet down for new and existing (retrofit) boiler equipment. If enacted, proposed Clear Skies legislation will, by 2008, require an average, effective, domestic NOx emissions rate of 0.16 lb/MMBtu, which number will be reduced to 0.13 lb/MMBtu by 2018. Such levels represent a 60% and 67% reduction, respectively, from the effective 2000 level of 0.40 lb/MMBtu. Low cost solutions to meet such regulations, and in particular those that can avoid the need for a costly selective catalytic reduction system (SCR), provide a strong incentive to continue to improve low NOx firing system technology to meet current and anticipated NOx control regulations. In light of these needs, ALSTOM, in cooperation with the DOE, is developing an enhanced combustion, low NOx pulverized coal burner which, when integrated with ALSTOM's state-of-the-art, globally air staged low NOx firing systems, will provide a means to achieve less than 0.15 lb/MMBtu NOx at less than 3/4 the cost of an SCR with low to no impact on balance of plant issues when firing a high volatile bituminous coal. Such coals can be more economic to fire than subbituminous or Powder River Basin (PRB) coals, but are more problematic from a NOx control standpoint as existing firing system technologies do not provide a means to meet current or anticipated regulations absent the use of an SCR. The DOE/ALSTOM program performed large pilot scale combustion testing in ALSTOM's Industrial Scale Burner Facility (ISBF) at its U.S. Power Plant Laboratories facility in Windsor, Connecticut. During this work, the near-field combustion environment was optimized to maximize NOx reduction while minimizing the impact on unburned carbon in ash, slagging and fouling, corrosion, and flame stability/turn-down under globally reducing conditions. Initially, ALSTOM utilized computational fluid dynamic modeling to evaluate a series of burner and/or near field stoichiometry controls in order to screen promising design concepts in advance of the large pilot scale testing. The third and final test, to be executed, will utilize several variants of the best nozzle tip configuration and compare performance with 3 different coals. The fuels to be tested will cover a wide range of coals commonly fired at US utilities. The completion of this work will provide sufficient data to allow ALSTOM to design, construct, and demonstrate a commercial version of an enhanced combustion low NOx pulverized coal burner. A preliminary cost/performance analysis of the developed enhanced combustion low NOx burner applied to ALSTOM's state-of-the-art TFS 2000 firing system was performed to show that the burner enhancements is a cost effective means to reduce NOx.

  10. argonne advanced research reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    and aerospace industry. ... enables methods, is used for in-service inspection of nuclear power plant components, such as tubing, piping the safe operationof advanced nuclear...

  11. Advances in process intensification through multifunctional reactor engineering.

    SciTech Connect (OSTI)

    Cooper, Marcia A.; Miller, James Edward; O'Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01T23:59:59.000Z

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  12. ADVANCED REACTOR SAFETY PROGRAM – STAKEHOLDER INTERACTION AND FEEDBACK

    SciTech Connect (OSTI)

    Spencer, Benjamin W; Huang, Hai

    2014-08-01T23:59:59.000Z

    In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  13. Uniform-burning matrix burner

    DOE Patents [OSTI]

    Bohn, Mark S. (Golden, CO); Anselmo, Mark (Arvada, CO)

    2001-01-01T23:59:59.000Z

    Computer simulation was used in the development of an inward-burning, radial matrix gas burner and heat pipe heat exchanger. The burner and exchanger can be used to heat a Stirling engine on cloudy days when a solar dish, the normal source of heat, cannot be used. Geometrical requirements of the application forced the use of the inward burning approach, which presents difficulty in achieving a good flow distribution and air/fuel mixing. The present invention solved the problem by providing a plenum with just the right properties, which include good flow distribution and good air/fuel mixing with minimum residence time. CFD simulations were also used to help design the primary heat exchanger needed for this application which includes a plurality of pins emanating from the heat pipe. The system uses multiple inlet ports, an extended distance from the fuel inlet to the burner matrix, flow divider vanes, and a ring-shaped, porous grid to obtain a high-temperature uniform-heat radial burner. Ideal applications include dish/Stirling engines, steam reforming of hydrocarbons, glass working, and any process requiring high temperature heating of the outside surface of a cylindrical surface.

  14. advanced test idaho reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    test idaho reactor First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 DISPERSAL AND HARVEST OF SAGE GROUSE...

  15. Advances in the development of wire mesh reactor for coal gasification studies - article no. 084102

    SciTech Connect (OSTI)

    Zeng, C.; Chen, L.; Liu, G.; Li, W.H.; Huang, B.M.; Zhu, H.D.; Zhang, B.; Zamansky, V. [GE Global Research Shanghai, Shanghai (China)

    2008-08-15T23:59:59.000Z

    In an effort to further understand the coal gasification behavior in entrained-flow gasifiers, a high pressure and high temperature wire mesh reactor with new features was recently built. An advanced LABVIEW-based temperature measurement and control system were adapted. Molybdenum wire mesh with aperture smaller than 70 {mu} m and type D thermocouple were used to enable high carbon conversion ({gt}90%) at temperatures {gt}1000 {sup o}C. Gaseous species from wire mesh reactor were quantified using a high sensitivity gas chromatography. The material balance of coal pyrolysis in wire mesh reactor was demonstrated for the first time by improving the volatile's quantification techniques.

  16. Status of development and licensing support for advanced liquid metal reactors in the United States

    SciTech Connect (OSTI)

    Pedersen, D.R. (Argonne National Lab., IL (United States)); Gyorey, G. (General Electric Co., San Jose, CA (United States))

    1991-01-01T23:59:59.000Z

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment.

  17. Safety aspects of the US advanced LMR (liquid metal reactor) design

    SciTech Connect (OSTI)

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S. (Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA); Argonne National Lab., IL (USA); USDOE Assistant Secretary for Nuclear Energy, Washington, DC (USA))

    1989-01-01T23:59:59.000Z

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs.

  18. Status of development and licensing support for advanced liquid metal reactors in the United States

    SciTech Connect (OSTI)

    Pedersen, D.R. [Argonne National Lab., IL (United States); Gyorey, G. [General Electric Co., San Jose, CA (United States)

    1991-12-01T23:59:59.000Z

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment.

  19. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01T23:59:59.000Z

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

  20. advanced reactors part: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 41 NC Smart Fleet Initiative is supported in part through the Clean Fuels Advanced Technology...

  1. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect (OSTI)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01T23:59:59.000Z

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  2. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect (OSTI)

    Not Listed

    2011-11-01T23:59:59.000Z

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  3. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    SciTech Connect (OSTI)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01T23:59:59.000Z

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  4. Comments on: Consortium for Advanced Simulation of Light Water Reactors

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i tCollaboration

  5. Advanced reactor safety research. Quarterly report, July-September 1982. Volume 23

    SciTech Connect (OSTI)

    Not Available

    1984-01-01T23:59:59.000Z

    Information is presented concerning core debris behavior; high-temperature fission-product chemistry and transport; containment analysis; elevated temperature materials assessment; development of LMFBR regulatory criteria and source terms; advanced reactor core phenomenology; LWR damaged fuel phenomenology; and ACRR status.

  6. 2012 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  7. 2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  8. 2010 Radiological Monitoring Results Associated with the Advance Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    mike lewis

    2011-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  9. 2011 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2012-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  10. Advanced Computational Thermal Studies and their Assessment for Supercritical-Pressure Reactors (SCRs)

    SciTech Connect (OSTI)

    D. M. McEligot; J. Y. Yoo; J. S. Lee; S. T. Ro; E. Lurien; S. O. Park; R. H. Pletcher; B. L. Smith; P. Vukoslavcevic; J. M. Wallace

    2009-04-01T23:59:59.000Z

    The goal of this laboratory / university collaboration of coupled computational and experimental studies is the improvement of predictive methods for supercritical-pressure reactors. The general objective is to develop supporting knowledge needed of advanced computational techniques for the technology development of the concepts and their safety systems.

  11. Advanced reactors transition fiscal year 1996 multi-year program plan WBS 7.3

    SciTech Connect (OSTI)

    Hulvey, R.K.

    1995-09-21T23:59:59.000Z

    This document describes in detail the work to be accomplished in FY 1996 and the out years for the Advanced Reactors Transition (WBS 7.3. ). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1996 Multi-Year Program Plan (MYPP), DOE will provide authorization to perform the work outlined in the FY 1996 MYPP

  12. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect (OSTI)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01T23:59:59.000Z

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  13. REQUEST BY SIEMENS WESTINGHOUSE POWER CORPORATION FOR AN ADVANCE...

    Broader source: Energy.gov (indexed) [DOE]

    topping combustor are called the multi-annular swirl burner (MASB) and the Piloted Syngas Burner (PSB). The work is sponsored by the Office of Fossil Energy. An advance waiver...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR Reactor Vessel

  15. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR Reactor VesselRadiation

  16. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR Reactor

  17. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR ReactorIndustry Council and

  18. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR ReactorIndustry Council

  19. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR ReactorIndustry

  20. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR ReactorIndustryMedia Center

  1. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting theCommercializationValidation and UncertaintyPWR ReactorIndustryMedia

  2. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect (OSTI)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01T23:59:59.000Z

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  3. Development of a Low Pressure, Air Atomized Oil Burner with High Atomizer Air Flow: Progress Report FY 1997

    SciTech Connect (OSTI)

    Butcher, T.A.

    1998-01-01T23:59:59.000Z

    This report describes technical advances made to the concept of a low pressure, air atomized oil burner for home heating applications. Currently all oil burners on the market are of the pressure atomized, retention head type. These burners have a lower firing rate limit of about 0.5 gallons per hour of oil, due to reliability problems related to small flow passage sizes. High pressure air atomized burners have been shown to be one route to avoid this problem but air compressor cost and reliability have practically eliminated this approach. With the low pressure air atomized burner the air required for atomization can be provided by a fan at 5-8 inches of water pressure. A burner using this concept, termed the Fan-Atomized Burner or ''FAB'' has been developed and is currently being commercialized. In the head of the FAB, the combustion air is divided into three parts, much like a conventional retention head burner. This report describes development work on a new concept in which 100% of the air from the fan goes through the atomizer. The primary advantage of this approach is a great simplification of the head design. A nozzle specifically sized for this concept was built and is described in the report. Basic flow pressure tests, cold air velocity profiles, and atomization performance have been measured. A burner head/flame tube has been developed which promotes a toroidal recirculation zone near the nozzle for flame stability. The burner head has been tested in several furnace and boiler applications over the firing rate range 0.2 to 0.28 gallons per hour. In all cases the burner can operate with very low excess air levels (under 10%) without producing smoke. Flue gas NO{sub x} concentration varied from 42 to 62 ppm at 3% O{sub 2}. The concept is seen as having significant potential and planned development efforts are discussed.

  4. Coleman Two Burner Stove The Coleman Matchlight 2-Burner Propane Stove is especially designed for outdoor

    E-Print Network [OSTI]

    Walker, Lawrence R.

    Coleman Two Burner Stove The Coleman Matchlight 2-Burner Propane Stove is especially designed-burner propane stove has a high-pressure regulator that ensures a constant flame regardless of weather propane stove has a removable nickel-chrome-plated grate that makes for easy cleaning. The aluminized

  5. Advance Reactor Concepts Technical Review Panel Public Report | Department

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed off Energy.gov. Are you0 ARRA Newsletters 20103-03 AUDITProductsletter No.10-006 Advance Patentof

  6. Advanced Reactor Concepts Technical Review Panel Report | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny: The Future of1 AAcceleratedDepartment ofDepartment ofMachinesEnergy Advanced

  7. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. B. Grover

    2007-05-01T23:59:59.000Z

    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 during irradiation were completed in 2006. The experiment was inserted in the ATR in December 2006, and will serve 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.

  8. Combined Heat and Power (CHP) Integrated with Burners for Packaged...

    Broader source: Energy.gov (indexed) [DOE]

    Combined Heat and Power (CHP) Integrated with Burners for Packaged Boilers Combined Heat and Power (CHP) Integrated with Burners for Packaged Boilers Providing Clean, Low-Cost,...

  9. A Novel Approach to Materials Development for Advanced Reactor Systems. Annual Report for Year 1

    SciTech Connect (OSTI)

    Was, G.S.; Atzmon, M.; Wang, L.

    2000-09-28T23:59:59.000Z

    OAK B188 A Novel Approach to Materials Development for Advanced Reactor Systems. Annual Report for Year 1 Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use of chromium pre-enrichment and the combination of cold-work and irradiation hardening in an effort to assess the role of radiation damage in IASCC without the effects of RIS. Third, to initiate irradiation of reactor pressure vessel steel and Zircaloy. Program Achievements for Year One: Progress was made on all 4 tasks in year one.

  10. DEVELOPMENT OF HUMAN FACTORS ENGINEERING GUIDANCE FOR SAFETY EVALUATIONS OF ADVANCED REACTORS.

    SciTech Connect (OSTI)

    O'HARA, J.; PERSENSKY, J.; SZABO, A.

    2006-10-01T23:59:59.000Z

    Advanced reactors are expected to be based on a concept of operations that is different from what is currently used in today's reactors. Therefore, regulatory staff may need new tools, developed from the best available technical bases, to support licensing evaluations. The areas in which new review guidance may be needed and the efforts underway to address the needs will be discussed. Our preliminary results focus on some of the technical issues to be addressed in three areas for which new guidance may be developed: automation and control, operations under degraded conditions, and new human factors engineering methods and tools.

  11. Secondary heat exchanger design and comparison for advanced high temperature reactor

    SciTech Connect (OSTI)

    Sabharwall, P. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States); Kim, E. S. [Seoul National Univ., P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Siahpush, A.; McKellar, M.; Patterson, M. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States)

    2012-07-01T23:59:59.000Z

    Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

  12. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

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

  13. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01T23:59:59.000Z

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.

  14. A review of two recent occurrences at the Advanced Test Reactor involving subcontractor activities

    SciTech Connect (OSTI)

    Dahlke, H.J.; Jensen, N.C.; Vail, J.A.

    1997-11-01T23:59:59.000Z

    This report documents the results of a brief, unofficial investigation into two incidents at the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR) facility, reported on October 25 and 31, 1997. The first event was an unanticipated breach of confinement. The second involved reactor operation with an inoperable seismic scram subsystem, violating the reactor`s Technical Specifications. These two incidents have been found to be unrelated. A third event that occurred on December 16, 1996, is also discussed because of its similarities to the first event listed above. Both of these incidents were unanticipated breaches of confinement, and both involved the work of construction subcontractor personnel. The cause for the subcontractor related occurrences is a work control process that fails to effectively interface with LMITCO management. ATR Construction Project managers work sufficient close with construction subcontractor personnel to understand planned day-to-day activities. They also have sufficient training and understanding of reactor operations to ensure adherence to applicable administrative requirements. However, they may not be sufficiently involved in the work authorization and control process to bridge an apparent communications gap between subcontractor employees and Facility Operations/functional support personnel for work inside the reactor facility. The cause for the inoperable seismic scram switch (resulting from a disconnected lead) is still under investigation. It does not appear to be subcontractor related.

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

    SciTech Connect (OSTI)

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

    2001-08-20T23:59:59.000Z

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

  16. Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    SciTech Connect (OSTI)

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

    1994-07-01T23:59:59.000Z

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES&H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power.

  17. Porous radiant burners having increased radiant output

    DOE Patents [OSTI]

    Tong, Timothy W. (Tempe, AZ); Sathe, Sanjeev B. (Tempe, AZ); Peck, Robert E. (Tempe, AZ)

    1990-01-01T23:59:59.000Z

    Means and methods for enhancing the output of radiant energy from a porous radiant burner by minimizing the scattering and increasing the adsorption, and thus emission of such energy by the use of randomly dispersed ceramic fibers of sub-micron diameter in the fabrication of ceramic fiber matrix burners and for use therein.

  18. ADVANCED COMPUTATIONAL MODEL FOR THREE-PHASE SLURRY REACTORS

    SciTech Connect (OSTI)

    Goodarz Ahmadi

    2001-10-01T23:59:59.000Z

    In the second year of the project, the Eulerian-Lagrangian formulation for analyzing three-phase slurry flows in a bubble column is further developed. The approach uses an Eulerian analysis of liquid flows in the bubble column, and makes use of the Lagrangian trajectory analysis for the bubbles and particle motions. An experimental set for studying a two-dimensional bubble column is also developed. The operation of the bubble column is being tested and diagnostic methodology for quantitative measurements is being developed. An Eulerian computational model for the flow condition in the two-dimensional bubble column is also being developed. The liquid and bubble motions are being analyzed and the results are being compared with the experimental setup. Solid-fluid mixture flows in ducts and passages at different angle of orientations were analyzed. The model predictions were compared with the experimental data and good agreement was found. Gravity chute flows of solid-liquid mixtures is also being studied. Further progress was also made in developing a thermodynamically consistent model for multiphase slurry flows with and without chemical reaction in a state of turbulent motion. The balance laws are obtained and the constitutive laws are being developed. Progress was also made in measuring concentration and velocity of particles of different sizes near a wall in a duct flow. The technique of Phase-Doppler anemometry was used in these studies. The general objective of this project is to provide the needed fundamental understanding of three-phase slurry reactors in Fischer-Tropsch (F-T) liquid fuel synthesis. The other main goal is to develop a computational capability for predicting the transport and processing of three-phase coal slurries. The specific objectives are: (1) To develop a thermodynamically consistent rate-dependent anisotropic model for multiphase slurry flows with and without chemical reaction for application to coal liquefaction. Also establish the material parameters of the model. (2) To provide experimental data for phasic fluctuation and mean velocities, as well as the solid volume fraction in the shear flow devices. (3) To develop an accurate computational capability incorporating the new rate-dependent and anisotropic model for analyzing reacting and nonreacting slurry flows, and to solve a number of technologically important problems related to Fischer-Tropsch (F-T) liquid fuel production processes. (4) To verify the validity of the developed model by comparing the predicted results with the performed and the available experimental data under idealized conditions.

  19. ADVANCED COMPUTATIONAL MODEL FOR THREE-PHASE SLURRY REACTORS

    SciTech Connect (OSTI)

    Goodarz Ahmadi

    2000-11-01T23:59:59.000Z

    In the first year of the project, solid-fluid mixture flows in ducts and passages at different angle of orientations were analyzed. The model predictions are compared with the experimental data and good agreement was found. Progress was also made in analyzing the gravity chute flows of solid-liquid mixtures. An Eulerian-Lagrangian formulation for analyzing three-phase slurry flows in a bubble column is being developed. The approach uses an Eulerian analysis of gas liquid flows in the bubble column, and makes use of the Lagrangian particle tracking procedure to analyze the particle motions. Progress was also made in developing a rate dependent thermodynamically consistent model for multiphase slurry flows in a state of turbulent motion. The new model includes the effect of phasic interactions and leads to anisotropic effective phasic stress tensors. Progress was also made in measuring concentration and velocity of particles of different sizes near a wall in a duct flow. The formulation of a thermodynamically consistent model for chemically active multiphase solid-fluid flows in a turbulent state of motion was also initiated. The general objective of this project is to provide the needed fundamental understanding of three-phase slurry reactors in Fischer-Tropsch (F-T) liquid fuel synthesis. The other main goal is to develop a computational capability for predicting the transport and processing of three-phase coal slurries. The specific objectives are: (1) To develop a thermodynamically consistent rate-dependent anisotropic model for multiphase slurry flows with and without chemical reaction for application to coal liquefaction. Also to establish the material parameters of the model. (2) To provide experimental data for phasic fluctuation and mean velocities, as well as the solid volume fraction in the shear flow devices. (3) To develop an accurate computational capability incorporating the new rate-dependent and anisotropic model for analyzing reacting and nonreacting slurry flows, and to solve a number of technologically important problems related to Fischer-Tropsch (F-T) liquid fuel production processes. (4) To verify the validity of the developed model by comparing the predicted results with the performed and the available experimental data under idealized conditions.

  20. In-Situ Creep Testing Capability for the Advanced Test Reactor

    SciTech Connect (OSTI)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2012-09-01T23:59:59.000Z

    An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

  1. Modeling & analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    SciTech Connect (OSTI)

    Georgevich, V.; Kim, S.H.; Taleyarkhan, R.P.; Jackson, S.

    1992-10-01T23:59:59.000Z

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied.

  2. Apollo - An advanced fuel fusion power reactor for the 21st century

    SciTech Connect (OSTI)

    Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

    1989-03-01T23:59:59.000Z

    A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enables the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor.

  3. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    SciTech Connect (OSTI)

    Wood, RT

    2004-09-27T23:59:59.000Z

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  4. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    SciTech Connect (OSTI)

    Kiselyov, V.A.; Sokov, L.M.

    1997-04-01T23:59:59.000Z

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

  5. Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

    SciTech Connect (OSTI)

    William Richins; Stephen Novascone; Cheryl O'Brien

    2009-08-01T23:59:59.000Z

    Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  6. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect (OSTI)

    C. A. Wemple

    2005-09-01T23:59:59.000Z

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  7. System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Craig Wise

    2011-12-01T23:59:59.000Z

    Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

  8. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect (OSTI)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01T23:59:59.000Z

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  9. Diesel fuel burner for diesel emissions control system

    DOE Patents [OSTI]

    Webb, Cynthia C.; Mathis, Jeffrey A.

    2006-04-25T23:59:59.000Z

    A burner for use in the emissions system of a lean burn internal combustion engine. The burner has a special burner head that enhances atomization of the burner fuel. Its combustion chamber is designed to be submersed in the engine exhaust line so that engine exhaust flows over the outer surface of the combustion chamber, thereby providing efficient heat transfer.

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

    SciTech Connect (OSTI)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01T23:59:59.000Z

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

  11. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    SciTech Connect (OSTI)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24T23:59:59.000Z

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well-suited to coupling with the unstructured meshes that are used in other physics simulations.

  12. The second and third NGNP advanced gas reactor fuel irradiation experiments

    SciTech Connect (OSTI)

    Grover, S. B.; Petti, D. A. [Idaho National Laboratory, 2525 N. Fremont Ave., Idaho Falls, ID 83415 (United States)

    2012-07-01T23:59:59.000Z

    The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

  13. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    SciTech Connect (OSTI)

    Donna P. Guillen

    2012-07-01T23:59:59.000Z

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  14. The benefits of an advanced fast reactor fuel cycle for plutonium management

    SciTech Connect (OSTI)

    Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

    1996-12-31T23:59:59.000Z

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.

  15. Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

    SciTech Connect (OSTI)

    J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

    2011-05-31T23:59:59.000Z

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The contributions of SCSU are described in Section 4, including a summary of undergraduate research exper

  16. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01T23:59:59.000Z

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

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

    SciTech Connect (OSTI)

    Ingersoll, D.T.

    2004-07-29T23:59:59.000Z

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.

  18. Development of Regulatory Technical Requirements for the Advanced Integral Type Research Reactor

    SciTech Connect (OSTI)

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik; Kim, Hho Jung [Korea Institute of Nuclear Safety, 19 Kusung-dong, Yusung-ku, Taejon, 305-338 (Korea, Republic of)

    2004-07-01T23:59:59.000Z

    This paper presents the current status of the study on the development of regulatory technical requirements for the licensing review of an advanced integral type research reactor of which the license application is expected in a few years. According to the Atomic Energy Act of Korea, both research and education reactors are subject to the technical requirements for power reactors in the licensing review. But, some of the requirements may not be applicable or insufficient for the licensing reviews of reactors with unique design features. Thus it is necessary to identify which review topics or areas can not be addressed by the existing requirements and to develop the required ones newly or supplement appropriately. Through the study performed so far, it has been identified that the following requirements need to be developed newly for the licensing review of SMART-P: the use of proven technology, the interfacial facility, the non-safety systems, and the metallic fuels. The approach and basis for the development of each of the requirements are discussed. (authors)

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

    SciTech Connect (OSTI)

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

    2002-01-01T23:59:59.000Z

    The focus of this program is to provide insight into the formation and minimization of NO{sub x} in multi-burner arrays, such as those that would be found in a typical utility boiler. Most detailed studies are performed in single-burner test facilities, and may not capture significant burner-to-burner interactions that could influence NO{sub x} emissions. Thus, investigations of such interactions were made by performing a combination of single and multiple burner experiments in a pilot-scale coal-fired test facility at the University of Utah, and by the use of computational combustion simulations to evaluate full-scale utility boilers. In addition, fundamental studies on nitrogen release from coal were performed to develop greater understanding of the physical processes that control NO formation in pulverized coal flames--particularly under low NO{sub x} conditions. A CO/H{sub 2}/O{sub 2}/N{sub 2} flame was operated under fuel-rich conditions in a flat flame reactor to provide a high temperature, oxygen-free post-flame environment to study secondary reactions of coal volatiles. Effects of temperature, residence time and coal rank on nitrogen evolution and soot formation were examined. Elemental compositions of the char, tar and soot were determined by elemental analysis, gas species distributions were determined using FTIR, and the chemical structure of the tar and soot was analyzed by solid-state {sup 13}C NMR spectroscopy. A laminar flow drop tube furnace was used to study char nitrogen conversion to NO. The experimental evidence and simulation results indicated that some of the nitrogen present in the char is converted to nitric oxide after direct attack of oxygen on the particle, while another portion of the nitrogen, present in more labile functionalities, is released as HCN and further reacts in the bulk gas. The reaction of HCN with NO in the bulk gas has a strong influence on the overall conversion of char-nitrogen to nitric oxide; therefore, any model that aims to predict the conversion of char-nitrogen to nitric oxide should allow for the conversion of char-nitrogen to HCN. The extent of the HCN conversion to NO or N{sub 2} will depend on the composition of the atmosphere surrounding the particle. A pilot-scale testing campaign was carried out to evaluate the impact of multiburner firing on NO{sub x} emissions using a three-burner vertical array. In general, the results indicated that multiburner firing yielded higher NO{sub x} emissions than single burner firing at the same fuel rate and excess air. Mismatched burner operation, due to increases in the firing rate of the middle burner, generally demonstrated an increase in NO{sub x} over uniform firing. Biased firing, operating the middle burner fuel rich with the upper and lower burners fuel lean, demonstrated an overall reduction in NO{sub x} emissions; particularly when the middle burner was operated highly fuel rich. Computational modeling indicated that operating the three burner array with the center burner swirl in a direction opposite to the other two resulted in a slight reduction in NO{sub x}.

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

    SciTech Connect (OSTI)

    Not Available

    1980-12-01T23:59:59.000Z

    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.

  1. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01T23:59:59.000Z

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  2. Fabrication of advanced oxide fuels containing minor actinide for use in fast reactors

    SciTech Connect (OSTI)

    Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Oarai-machi, Higashi-ibaraki-gun, Ibaraki, 311-1393 (Japan)

    2007-07-01T23:59:59.000Z

    R and D of advanced fuel containing minor actinide for use in fast reactors is described related to the composite fuel with MgO matrix. Fabrication tests of MgO composite fuels containing Am were done by a practical process that could be adapted to the presently used commercial manufacturing technology. Am-containing MgO composite fuels having good characteristics, i.e., having no defects, a high density, a homogeneous dispersion of host phase, were obtained. As related technology, burn-up characteristics of a fast reactor core loaded with the present MgO composite fuel were also analyzed, mainly in terms of core criticality. Furthermore, phase relations of MA oxide which was assumed to be contained in MgO matrix fuel were experimentally investigated. (authors)

  3. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    SciTech Connect (OSTI)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09T23:59:59.000Z

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  4. Reverberatory screen for a radiant burner

    DOE Patents [OSTI]

    Gray, Paul E. (North East, MD)

    1999-01-01T23:59:59.000Z

    The present invention relates to porous mat gas fired radiant burner panels utilizing improved reverberatory screens. The purpose of these screens is to boost the overall radiant output of the burner relative to a burner using no screen and the same fuel-air flow rates. In one embodiment, the reverberatory screen is fabricated from ceramic composite material, which can withstand higher operating temperatures than its metallic equivalent. In another embodiment the reverberatory screen is corrugated. The corrugations add stiffness which helps to resist creep and thermally induced distortions due to temperature or thermal expansion coefficient differences. As an added benefit, it has been unexpectedly discovered that the corrugations further increase the radiant efficiency of the burner. In a preferred embodiment, the reverberatory screen is both corrugated and made from ceramic composite material.

  5. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01T23:59:59.000Z

    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. 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. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilitiesmore »while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. 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, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.« less

  6. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01T23:59:59.000Z

    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. 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. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilities while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. 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, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.

  7. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2014-05-01T23:59:59.000Z

    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 undergo on-line fission product monitoring 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 sup

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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.

  9. Silane-propane ignitor/burner

    DOE Patents [OSTI]

    Hill, Richard W. (Livermore, CA); Skinner, Dewey F. (Livermore, CA); Thorsness, Charles B. (Livermore, CA)

    1985-01-01T23:59:59.000Z

    A silane propane burner for an underground coal gasification process which is used to ignite the coal and to controllably retract the injection point by cutting the injection pipe. A narrow tube with a burner tip is positioned in the injection pipe through which an oxidant (oxygen or air) is flowed. A charge of silane followed by a supply of fuel, such as propane, is flowed through the tube. The silane spontaneously ignites on contact with oxygen and burns the propane fuel.

  10. Silane-propane ignitor/burner

    DOE Patents [OSTI]

    Hill, R.W.; Skinner, D.F. Jr.; Thorsness, C.B.

    1983-05-26T23:59:59.000Z

    A silane propane burner for an underground coal gasification process which is used to ignite the coal and to controllably retract the injection point by cutting the injection pipe. A narrow tube with a burner tip is positioned in the injection pipe through which an oxidant (oxygen or air) is flowed. A charge of silane followed by a supply of fuel, such as propane, is flowed through the tube. The silane spontaneously ignites on contact with oxygen and burns the propane fuel.

  11. Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently20,000 Russian NuclearandJune 17, 2015EnergyTheAdvancedReactorEnergy Technology |

  12. Burners and combustion apparatus for carbon nanomaterial production

    DOE Patents [OSTI]

    Alford, J. Michael (Lakewood, CO); Diener, Michael D. (Denver, CO); Nabity, James (Arvada, CO); Karpuk, Michael (Boulder, CO)

    2007-10-09T23:59:59.000Z

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

  13. Burners and combustion apparatus for carbon nanomaterial production

    DOE Patents [OSTI]

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

    2013-02-05T23:59:59.000Z

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

  14. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  15. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

    SciTech Connect (OSTI)

    Curtis Smith

    2013-09-01T23:59:59.000Z

    During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

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

    SciTech Connect (OSTI)

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

    2007-09-01T23:59:59.000Z

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  17. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect (OSTI)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01T23:59:59.000Z

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  18. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    SciTech Connect (OSTI)

    Donna Post Guillen

    2012-11-01T23:59:59.000Z

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  19. The Advanced Test Reactor Irradiation Capabilities Available as a National Scientific User Facility

    SciTech Connect (OSTI)

    S. Blaine Grover

    2008-09-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

  20. Improved computational neutronics methods and validation protocols for the advanced test reactor

    SciTech Connect (OSTI)

    Nigg, D. W.; Nielsen, J. W.; Chase, B. M.; Murray, R. K.; Steuhm, K. A.; Unruh, T. [Idaho National Laboratory, 2525 Fremont Street, Idaho Falls, ID 83415-3870 (United States)

    2012-07-01T23:59:59.000Z

    The Idaho National Laboratory (INL) is in the process of updating the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purposes. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry have been conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for flexible and repeatable ATR physics code validation protocols that are consistent with applicable national standards. (authors)

  1. Measurements of the subcriticality using advanced technique of shooting source during operation of NPP reactors

    SciTech Connect (OSTI)

    Lebedev, G. V., E-mail: lgv2004@mail.ru; Petrov, V. V. [National Research Center Kurchatov Institute (Russian Federation); Bobylyov, V. T.; Butov, R. I.; Zhukov, A. M.; Sladkov, A. A. [Dukhov VNIIA (Russian Federation)

    2014-12-15T23:59:59.000Z

    According to the rules of nuclear safety, the measurements of the subcriticality of reactors should be carried out in the process of performing nuclear hazardous operations. An advanced technique of shooting source of neutrons is proposed to meet this requirement. As such a source, a pulsed neutron source (PNS) is used. In order to realize this technique, it is recommended to enable a PNS with a frequency of 1–20 Hz. The PNS is stopped after achieving a steady-state (on average) number of neutrons in the reactor volume. The change in the number of neutrons in the reactor volume is measured in time with an interval of discreteness of ?0.1 s. The results of these measurements with the application of a system of point-kinetics equations are used in order to calculate the sought subcriticality. The basic idea of the proposed technique used to measure the subcriticality is elaborated in a series of experiments on the Kvant assembly. The conditions which should be implemented in order to obtain a positive result of measurements are formulated. A block diagram of the basic version of the experimental setup is presented, whose main element is a pulsed neutron generator.

  2. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect (OSTI)

    Kim, T. K.; Grandy, C.; Hill, R. N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01T23:59:59.000Z

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

  3. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01T23:59:59.000Z

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  4. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01T23:59:59.000Z

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  5. Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Yoon, Su-Jong [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Kim, Eung-Soo [Seoul National Univ., Seoul (Korea, Republic of)

    2014-03-01T23:59:59.000Z

    Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution of single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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 significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

  7. Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

    2013-04-04T23:59:59.000Z

    Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different requirements, including the need to operate in different coolant environments, higher operating temperatures, and longer operating cycles between planned refueling and maintenance outages. These features, along with the relative lack of operating experience for some of the proposed advanced designs, may limit the ability to estimate event probability and component POF with a high degree of certainty. Incorporating real-time estimates of component POF may compensate for a relative lack of established knowledge about the long-term component behavior and improve operational and maintenance planning and optimization. The particular eccentricities of advanced reactors and small modular reactors provide unique challenges and needs for advanced instrumentation, control, and human-machine interface (ICHMI) techniques such as enhanced risk monitors (ERM) in aSMRs. Several features of aSMR designs increase the need for accurate characterization of the real-time risk during operation and maintenance activities. A number of technical gaps in realizing ERM exist, and these gaps are largely independent of the specific reactor technology. As a result, the development of a framework for ERM would enable greater situational awareness regardless of the specific class of reactor technology. A set of research tasks are identified in a preliminary research plan to enable the development, testing, and demonstration of such a framework. Although some aspects of aSMRs, such as specific operational characteristics, will vary and are not now completely defined, the proposed framework is expected to be relevant regardless of such uncertainty. The development of an ERM framework will provide one of the key technical developments necessary to ensure the economic viability of aSMRs.

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

    SciTech Connect (OSTI)

    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

    2014-02-01T23:59:59.000Z

    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 insertion into a commercial reactor within the desired timeframe (by 2022).

  9. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect (OSTI)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01T23:59:59.000Z

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  10. CHP Integrated with Burners for Packaged Boilers

    SciTech Connect (OSTI)

    Castaldini, Carlo; Darby, Eric

    2013-09-30T23:59:59.000Z

    The objective of this project was to engineer, design, fabricate, and field demonstrate a Boiler Burner Energy System Technology (BBEST) that integrates a low-cost, clean burning, gas-fired simple-cycle (unrecuperated) 100 kWe (net) microturbine (SCMT) with a new ultra low-NOx gas-fired burner (ULNB) into one compact Combined Heat and Power (CHP) product that can be retrofit on new and existing industrial and commercial boilers in place of conventional burners. The Scope of Work for this project was segmented into two principal phases: (Phase I) Hardware development, assembly and pre-test and (Phase II) Field installation and demonstration testing. Phase I was divided into five technical tasks (Task 2 to 6). These tasks covered the engineering, design, fabrication, testing and optimization of each key component of the CHP system principally, ULNB, SCMT, assembly BBEST CHP package, and integrated controls. Phase I work culminated with the laboratory testing of the completed BBEST assembly prior to shipment for field installation and demonstration. Phase II consisted of two remaining technical tasks (Task 7 and 8), which focused on the installation, startup, and field verification tests at a pre-selected industrial plant to document performance and attainment of all project objectives. Technical direction and administration was under the management of CMCE, Inc. Altex Technologies Corporation lead the design, assembly and testing of the system. Field demonstration was supported by Leva Energy, the commercialization firm founded by executives at CMCE and Altex. Leva Energy has applied for patent protection on the BBEST process under the trade name of Power Burner and holds the license for the burner currently used in the product. The commercial term Power Burner is used throughout this report to refer to the BBEST technology proposed for this project. The project was co-funded by the California Energy Commission and the Southern California Gas Company (SCG), a division of Sempra Energy. These match funds were provided via concurrent contracts and investments available via CMCE, Altex, and Leva Energy The project attained all its objectives and is considered a success. CMCE secured the support of GI&E from Italy to supply 100 kW Turbec T-100 microturbines for the project. One was purchased by the project’s subcontractor, Altex, and a second spare was purchased by CMCE under this project. The microturbines were then modified to convert from their original recuperated design to a simple cycle configuration. Replacement low-NOx silo combustors were designed and bench tested in order to achieve compliance with the California Air Resources Board (CARB) 2007 emission limits for NOx and CO when in CHP operation. The converted microturbine was then mated with a low NOx burner provided by Altex via an integration section that allowed flow control and heat recovery to minimize combustion blower requirements; manage burner turndown; and recover waste heat. A new fully integrated control system was designed and developed that allowed one-touch system operation in all three available modes of operation: (1) CHP with both microturbine and burner firing for boiler heat input greater than 2 MMBtu/hr; (2) burner head only (BHO) when the microturbine is under service; and (3) microturbine only when boiler heat input requirements fall below 2 MMBtu/hr. This capability resulted in a burner turndown performance of nearly 10/1, a key advantage for this technology over conventional low NOx burners. Key components were then assembled into a cabinet with additional support systems for generator cooling and fuel supply. System checkout and performance tests were performed in the laboratory. The assembled system and its support equipment were then shipped and installed at a host facility where final performance tests were conducted following efforts to secure fabrication, air, and operating permits. The installed power burner is now in commercial operation and has achieved all the performance goals.

  11. Development of Mechanistic Modeling Capabilities for Local Neutronically-Coupled Flow-Induced Instabilities in Advanced Water-Cooled Reactors

    SciTech Connect (OSTI)

    Michael Podowski

    2009-11-30T23:59:59.000Z

    The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.

  12. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    SciTech Connect (OSTI)

    Not Available

    1992-08-01T23:59:59.000Z

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  13. Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

    SciTech Connect (OSTI)

    Chan, Wai Kit, E-mail: kekyeung@ust.hk [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Joueet, Justine; Heng, Samuel; Yeung, King Lun [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Schrotter, Jean-Christophe [Water Research Center of Veolia, Anjou Recherche, Chemin de la Digue, BP 76. 78603, Maisons Laffitte, Cedex (France)

    2012-05-15T23:59:59.000Z

    An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.

  14. The use of zirconium hydride blankets in a minor actinide/thorium burner sodium-cooled reactor for void coefficient control with particular reference to UK's plutonium disposition problem

    E-Print Network [OSTI]

    Arias, Francisco J.; Parks, Geoffrey T.

    2015-04-21T23:59:59.000Z

    The use of zirconium hydride (Th–ZrH1.6) blankets in a thorium-fuelled sodium-cooled reactor for void reactivity control with particular reference to UK's plutonium disposition problem is proposed and considered. It is shown that, with the use...

  15. News Letter Institute of Advanced Energy, Kyoto University

    E-Print Network [OSTI]

    Takada, Shoji

    of advanced nuclear reactors like fast reactors or fusion reactors, where materials undergo intensive

  16. Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

    SciTech Connect (OSTI)

    Hambley, D.I. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2013-07-01T23:59:59.000Z

    The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

  17. Human factors engineering evaluation of the Advanced Test Reactor Control Room

    SciTech Connect (OSTI)

    Boone, M.P.; Banks, W.W.

    1980-12-01T23:59:59.000Z

    The information presented here represents preliminary findings related to an ongoing human engineering evaluation of the Advanced Test Reactor (ATR) Control Room. Although many of the problems examined in this report have been previously noted by ATR operations personnel, the systematic approach used in this investigation produced many new insights. While many violations of Human Engineering military standards (MIL-STD) are noted, and numerous recommendations made, the recommendations should be examined cautiously. The reason for our suggested caution lies in the fact that many ATR operators have well over 10-years experience in operating the controls, meters, etc. Hence, it is assumed adaptation to the existing system is quite developed and the introduction of hardware/control changes, even though the changes enhance the system, may cause short-term (or long-term, depending upon the amount of operator experience and training) adjustment problems for operators adapting to the new controls/meters and physical layout.

  18. Advanced Test Reactor Critical Facility safety analysis report five year currency review

    SciTech Connect (OSTI)

    Napper, P.R.; Carpenter, W.R.; Garner, R.W.

    1991-05-01T23:59:59.000Z

    By DOE-ID Order 5481.1A, a five year currency review is required of the Safety Analysis Reports of all ID or ID contractor operations having hazards of a type and magnitude not routinely encountered and/or accepted by the public. In keeping with this order, a currency review has been performed of the Advanced Test Reactor Critical Facility (ADTRC) Safety Analysis Report (SAR), Issue 003, 1990. The objectives of this currency review were to: evaluate the content, completeness, clarity of presentation and compliance with NRC Regulatory Guides and DOE Orders, etc., and evaluate the technical content of the SAR, particularly the Technical Specifications, and to evaluate the safety of continued operation of the ATRC. The reviewers concluded that although improvements may be needed in the overall content, clarity, and demonstration of compliance with current orders and regulations, the safety of the ATRC is in no way compromised and no unreviewed safety questions were identified. 6 figs., 3 tabs.

  19. Analysis of core-concrete interaction event with flooding for the Advanced Neutron Source reactor

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.

    1993-11-01T23:59:59.000Z

    This paper discusses salient aspects of the methodology, assumptions, and modeling of various features related to estimation of source terms from an accident involving a molten core-concrete interaction event (with and without flooding) in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for this postulated severe accident. Several design features (such as rupture disks) are examined to study containment response during this severe accident. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms, which are then used for studying off-site radiological consequences and health effects for the support of the Conceptual Safety Analysis Report for ANS. The results are also to be used to examine the effectiveness of subpile room flooding during this type of severe accident.

  20. Containment performance analyses for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.

    1992-10-01T23:59:59.000Z

    This paper discusses salient aspects of methodology, assumptions, and modeling of various features related to estimation of source terms from two conservatively scoped severe accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code.

  1. The progress of neutron texture diffractometer at China Advanced Research Reactor

    E-Print Network [OSTI]

    Li, MeiJuan; Liu, YunTao; Tian, GengFang; Gao, JianBo; Yu, ZhouXiang; Li, YuQing; Wu, LiQi; Yang, LinFeng; Sun, Kai; Wang, HongLi; Chen, DongFeng

    2015-01-01T23:59:59.000Z

    The first neutron texture diffractometer in China has been built at China Advanced Research Reactor due to the strong demands of texture measurement with neutrons from domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study the texture in the commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of Zr alloy plate are presented. The comparison of texture measurement among different neutron texture diffractometer of HIPPO at LANSCE, Kowari at ANSTO and neutron texture diffractometer at CARR illustrates the reliable performance of this texture diffractometer.

  2. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    SciTech Connect (OSTI)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01T23:59:59.000Z

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  3. Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

    2012-06-01T23:59:59.000Z

    The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

  4. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    SciTech Connect (OSTI)

    NONE

    1994-04-30T23:59:59.000Z

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  5. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    SciTech Connect (OSTI)

    David W. Nigg

    2013-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  6. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  7. A fission matrix based validation protocol for computed power distributions in the advanced test reactor

    SciTech Connect (OSTI)

    Nielsen, J. W. [Idaho National Laboratory, MS 3840, PO Box 1625, Idaho Falls, ID 83415 (United States); Nigg, D. W. [Idaho National Laboratory, MS 3860, PO Box 1625, Idaho Falls, ID 83415 (United States); LaPorta, A. W. [Idaho National Laboratory, MS 7136, PO Box 1625, Idaho Falls, ID 83415 (United States)

    2013-07-01T23:59:59.000Z

    The Idaho National Laboratory (INL) has been engaged in a significant multi year effort to modernize the computational reactor physics tools and validation procedures used to support operations of the Advanced Test Reactor (ATR) and its companion critical facility (ATRC). Several new protocols for validation of computed neutron flux distributions and spectra as well as for validation of computed fission power distributions, based on new experiments and well-recognized least-squares statistical analysis techniques, have been under development. In the case of power distributions, estimates of the a priori ATR-specific fuel element-to-element fission power correlation and covariance matrices are required for validation analysis. A practical method for generating these matrices using the element-to-element fission matrix is presented, along with a high-order scheme for estimating the underlying fission matrix itself. The proposed methodology is illustrated using the MCNP5 neutron transport code for the required neutronics calculations. The general approach is readily adaptable for implementation using any multidimensional stochastic or deterministic transport code that offers the required level of spatial, angular, and energy resolution in the computed solution for the neutron flux and fission source. (authors)

  8. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    SciTech Connect (OSTI)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  9. Evaluation of current drive requirements and operating characteristics of a high bootstrap fraction advanced tokamak reactor

    SciTech Connect (OSTI)

    Houlberg, W.A.; Attenberger, S.E.

    1995-02-01T23:59:59.000Z

    The reactor potential of some advanced physics operating modes proposed for the TPX physics program are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor is analyzed because of its potential for steady-state, non-inductive operation with high bootstrap current fraction. Particle, energy and toroidal current equations are evolved to steady-state conditions using the 1-1/2-D time-dependent WHIST transport code. The solutions are therefore consistent with particle, energy and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80-90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q {equivalent_to} P{sub fus}/P{sub aux}, to variations in the plasma properties is examined. The auxiliary current drive power, P{sub aux} = P{sub LH} + P{sub FW}; bootstrap current fraction: current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10-30 in these calculations when overall Q {approx} 30.

  10. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    1] B. Farrar et. al. , Fast reactor decay heat removal:CA [2] B. Farrar et. al. , Fast reactor decay heat removal:They are 1) gas cooled fast reactors (GFR), 2) very high

  11. Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

    SciTech Connect (OSTI)

    Johanna Oxstrand; Katya Le Blanc

    2014-07-01T23:59:59.000Z

    The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

  12. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    systems for the Gas Cooled Fast Reactor (GCFR) includes theThey are 1) gas cooled fast reactors (GFR), 2) very high

  13. TEM Examination of Advanced Alloys Irradiated in ATR

    SciTech Connect (OSTI)

    Jian Gan, PhD

    2007-09-01T23:59:59.000Z

    Successful development of materials is critical to the deployment of advanced nuclear power systems. Irradiation studies of candidate materials play a vital role for better understanding materials performance under various irradiation environments of advanced system designs. In many cases, new classes of materials have to be investigated to meet the requirements of these advanced systems. For applications in the temperature range of 500 800ºC which is relevant to the fast neutron spectrum burner reactors for the Global Nuclear Energy Partnership (GNEP) program, oxide dispersion strengthened (ODS) and ferritic martensitic steels (e.g., MA957 and others) are candidates for advanced cladding materials. In the low temperature regions of the core (<600ºC), alloy 800H, HCM12A (also called T 122) and HT 9 have been considered.

  14. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    SciTech Connect (OSTI)

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

    2013-07-01T23:59:59.000Z

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)

  15. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    SciTech Connect (OSTI)

    Renae Soelberg

    2014-11-01T23:59:59.000Z

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at CAES by Boise State PI Janelle Wharry and Cory Dolph. PI Corey Dolph returned in early November to complete their research by performing nanoindentation on unirradiated specimens that will be used as a baseline for their research.

  16. Incorporating Equipment Condition Assessment in Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-10-01T23:59:59.000Z

    Advanced small modular reactors (aSMRs) can complement the current fleet of large light-water reactors in the USA for baseload and peak demand power production and process heat applications (e.g., water desalination, shale oil extraction, hydrogen production). The day-to-day costs of aSMRs are expected to be dominated by operations and maintenance (O&M); however, the effect of diverse operating missions and unit modularity on O&M is not fully understood. These costs could potentially be reduced by optimized scheduling, with risk-informed scheduling of maintenance, repair, and replacement of equipment. Currently, most nuclear power plants have a “living” probabilistic risk assessment (PRA), which reflects the as-operated, as-modified plant and combine event probabilities with population-based probability of failure (POF) for key components. “Risk monitors” extend the PRA by incorporating the actual and dynamic plant configuration (equipment availability, operating regime, environmental conditions, etc.) into risk assessment. In fact, PRAs are more integrated into plant management in today’s nuclear power plants than at any other time in the history of nuclear power. However, population-based POF curves are still used to populate fault trees; this approach neglects the time-varying condition of equipment that is relied on during standard and non-standard configurations. Equipment condition monitoring techniques can be used to estimate the component POF. Incorporating this unit-specific estimate of POF in the risk monitor can provide a more accurate estimate of risk in different operating and maintenance configurations. This enhanced risk assessment will be especially important for aSMRs that have advanced component designs, which don’t have an available operating history to draw from, and often use passive design features, which present challenges to PRA. This paper presents the requirements and technical gaps for developing a framework to integrate unit-specific estimates of POF into risk monitors, resulting in enhanced risk monitors that support optimized operation and maintenance of aSMRs.

  17. Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor

    SciTech Connect (OSTI)

    J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

    2006-10-01T23:59:59.000Z

    The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

  18. Advanced application of the discrete generalized multigroup method and recondensation to reactor analysis

    E-Print Network [OSTI]

    Everson, Matthew S

    2014-01-01T23:59:59.000Z

    Fine-group whole-core reactor analysis remains one of the long sought goals of the reactor physics community. Such a detailed analysis is typically too computationally expensive to be realized on anything except the largest ...

  19. Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16T23:59:59.000Z

    to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling...

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

    SciTech Connect (OSTI)

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

    2013-10-01T23:59:59.000Z

    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.

  1. Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems

    SciTech Connect (OSTI)

    Maslov,V.M.; Oblozinsky, P.; Herman, M.

    2008-12-01T23:59:59.000Z

    In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.

  2. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    SciTech Connect (OSTI)

    Dan Ogden

    2014-10-01T23:59:59.000Z

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

  3. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

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

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton

    2014-02-01T23:59:59.000Z

    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.

  5. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  6. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  7. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.

  8. Refinery burner simulation design architecture summary.

    SciTech Connect (OSTI)

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

    2011-10-01T23:59:59.000Z

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

  9. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. H.; Christensen, R.; Sabharwall, P.

    2015-07-01T23:59:59.000Z

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore »combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

  10. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect (OSTI)

    David Petti

    2014-06-01T23:59:59.000Z

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO2 particle fuel up to about 10% FIMA and 1150°C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250°C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150°C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.

  11. Safety Topic: Bunsen Burners and Hotplates

    E-Print Network [OSTI]

    Cohen, Robert E.

    a medium to medium-high setting of the hot plate to heat most liquids, including water. Do not use the high setting to heat low-boiling liquids. The hot plate surface can reach a maximum temperature of 540 °C · Do.med.cornell.edu/ehs/updates/bunsen_burner_safety.htm #12;Hot Plate Procedures · Use only heat-resistant, borosilicate glassware, and check for cracks

  12. PULSE DRYING EXPERIMENT AND BURNER CONSTRUCTION

    SciTech Connect (OSTI)

    Robert States

    2006-07-15T23:59:59.000Z

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

  13. Capabilities and Facilities Available at the Advanced Test Reactor to Support Development of the Next Generation Reactors

    SciTech Connect (OSTI)

    S. Blaine Grover; Raymond V. Furstenau

    2005-10-01T23:59:59.000Z

    The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. It is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The Irradiation Test Vehicle (ITV) installed in 1999 enhanced these capabilities by providing a built in experiment monitoring and control system for instrumented and/or temperature controlled experiments. This built in control system significantly reduces the cost for an actively monitored/temperature controlled experiments by providing the thermocouple connections, temperature control system, and temperature control gas supply and exhaust systems already in place at the irradiation position. Although the ITV in-core hardware was removed from the ATR during the last core replacement completed in early 2005, it (or a similar facility) could be re-installed for an irradiation program when the need arises. The proposed Gas Test Loop currently being designed for installation in the ATR will provide additional capability for testing of not only gas reactor materials and fuels but will also include enhanced fast flux rates for testing of materials and fuels for other next generation reactors including preliminary testing for fast reactor fuels and materials. This paper discusses the different irradiation capabilities available and the cost benefit issues related to each capability.

  14. Coal-water mixture fuel burner

    DOE Patents [OSTI]

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

    1985-04-29T23:59:59.000Z

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

  15. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect (OSTI)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01T23:59:59.000Z

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  16. actinide burner fuel: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    in Great Britain developed standards for register type burners installed in fossil fuel fired electric generating... Cawte, A. D. 1979-01-01 3 Chemical and toxicological...

  17. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01T23:59:59.000Z

    New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

  18. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01T23:59:59.000Z

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  19. advanced water-cooled reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  20. advanced gas-cooled reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  1. advanced light-water reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  2. advanced gas-cooled reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  3. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    SciTech Connect (OSTI)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01T23:59:59.000Z

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  4. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    SciTech Connect (OSTI)

    David W. Nigg; Devin A. Steuhm

    2011-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

  5. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17T23:59:59.000Z

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.

  6. advanced gas cooled graphite moderated reactor: Topics by E-print...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    temperatures during normal (more) Moore, Eugene James Thomas 2006-01-01 2 THORIUM FUEL CYCLES: A GRAPHITE-MODERATED MOLTEN SALT REACTOR Physics Websites Summary: ,...

  7. On the interest of carbon-coated plasma reactor for advanced gate stack etching processes

    SciTech Connect (OSTI)

    Ramos, R.; Cunge, G.; Joubert, O. [Freescale Semiconductor Inc., 850 Rue Jean Monnet, 38921 Crolles Cedex (France) and Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France); Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France)

    2007-03-15T23:59:59.000Z

    In integrated circuit fabrication the most wide spread strategy to achieve acceptable wafer-to-wafer reproducibility of the gate stack etching process is to dry-clean the plasma reactor walls between each wafer processed. However, inherent exposure of the reactor walls to fluorine-based plasma leads to formation and accumulation of nonvolatile fluoride residues (such as AlF{sub x}) on reactor wall surfaces, which in turn leads to process drifts and metallic contamination of wafers. To prevent this while keeping an Al{sub 2}O{sub 3} reactor wall material, a coating strategy must be used, in which the reactor is coated by a protective layer between wafers. It was shown recently that deposition of carbon-rich coating on the reactor walls allows improvements of process reproducibility and reactor wall protection. The authors show that this strategy results in a higher ion-to-neutral flux ratio to the wafer when compared to other strategies (clean or SiOCl{sub x}-coated reactors) because the carbon walls load reactive radical densities while keeping the same ion current. As a result, the etching rates are generally smaller in a carbon-coated reactor, but a highly anisotropic etching profile can be achieved in silicon and metal gates, whose etching is strongly ion assisted. Furthermore, thanks to the low density of Cl atoms in the carbon-coated reactor, silicon etching can be achieved almost without sidewall passivation layers, allowing fine critical dimension control to be achieved. In addition, it is shown that although the O atom density is also smaller in the carbon-coated reactor, the selectivity toward ultrathin gate oxides is not reduced dramatically. Furthermore, during metal gate etching over high-k dielectric, the low level of parasitic oxygen in the carbon-coated reactor also allows one to minimize bulk silicon reoxidation through HfO{sub 2} high-k gate dielectric. It is then shown that the BCl{sub 3} etching process of the HfO{sub 2} high-k material is highly selective toward the substrate in the carbon-coated reactor, and the carbon-coating strategy thus allows minimizing the silicon recess of the active area of transistors. The authors eventually demonstrate that the carbon-coating strategy drastically reduces on-wafer metallic contamination. Finally, the consumption of carbon from the reactor during the etching process is discussed (and thus the amount of initial deposit that is required to protect the reactor walls) together with the best way of cleaning the reactor after a silicon etching process.

  8. E-Print Network 3.0 - advanced reactor technology Sample Search...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Op-ed Letters... to iron out their differences over the site of the International Thermonuclear Experimental Reactor... . Russia sticks to its guns, backs France for fusion...

  9. Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

    2009-01-01T23:59:59.000Z

    This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

  10. Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application 

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16T23:59:59.000Z

    abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark...

  11. Turbine Burners: Flameholding in Accelerating Flow W. A. Sirignano1

    E-Print Network [OSTI]

    Liu, Feng

    1 Turbine Burners: Flameholding in Accelerating Flow W. A. Sirignano1 , D. Dunn-Rankin2 , F. Liu3 B, Irvine Abstract A review of turbine-burner research and some relevant background issues is presented. Previous work on thermal cycle analysis for augmentative combustion in the passages of the turbine

  12. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01T23:59:59.000Z

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  13. Process Modeling Phase I Summary Report for the Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect (OSTI)

    Pannala, Sreekanth [ORNL; Daw, C Stuart [ORNL; Boyalakuntla, Dhanunjay S [ORNL; FINNEY, Charles E A [ORNL

    2006-09-01T23:59:59.000Z

    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 University. Prior to the current scale-up activity, considerable effort has gone in to adapting the MFIX code to incorporate the unique features of fuel coating reactors and also in validating the resulting simulation features with experimental observations. Much of this work is documented in previous AGR reports and publications (Pannala et al., 2004, Pannala et al., 2005, Boyalakuntla et al., 2005a, Boyalakuntla et al., 2005b and Finney et al., 2005). As a result of the previous work described above, the ORNL coater model now has the capability for simulating full spatio-temporal details of the gas-particle hydrodynamics and gas-particle heat and mass transfer in the TRISO coater. This capability provides a great deal of information about many of the processes believed to control quality, but the model is not yet sufficiently developed to fully predict coating quality for any given coater design and/or set of operating conditions because the detailed chemical reaction kinetics needed to make the model fully predictive are not yet available. Nevertheless, the model at its current stage of development already provides the most comprehensive and detailed quantitative information available about gas flows, solid flows, temperatures, and species inside the coater during operation. This level of information ought to be highly useful in expediting the scale-up process (e.g., in correlating observations and minimizing the number of pilot-scale tests required). However, previous work had not yet demonstrated that the typical design and/or operating changes known to affect product quality at the lab scale could be clearly discriminated by the existing model. The Modeling Scale-Up Phase I work was initiated to produce such a demonstration, and two detailed examples are discussed in this report.

  14. A Novel Approach to Material Development for Advanced Reactor Systems. Quarterly progress report, Year 1 - Quarter 2

    SciTech Connect (OSTI)

    None

    2000-03-27T23:59:59.000Z

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Quarterly progress report, Year 1--Quarter 2. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steels without the effects of RIS. Third, to prepare for the irradiation of reactor pressure vessel steel and Zircaloy. Program goals for Second Quarter, Year One: In year 1 quarter 2, the project goal was to complete an irradiation of an RPV steel sample and begin sample characterization. We also planned to identify sources of Zircaloy for irradiation and characterization.

  15. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    SciTech Connect (OSTI)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01T23:59:59.000Z

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390°C for KF ZrF4). The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses research efforts on the near-term qualification, selection, or maturation strategy as detailed in this report. Development of the integration methodology feasibility study, along with research and development (R&D) needs, are ongoing tasks that will be covered in the future reports as work progresses. Section 2 briefly presents the integration of AHTR technology with conventional chemical industrial processes., See Idaho National Laboratory (INL) TEV-1160 (2011) for further details

  16. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01T23:59:59.000Z

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten-rhenium and platinum rhodium thermocouples can be avoided. INL is also developing an Ultrasonic Thermometry (UT) capability. In addition to small size, UT’s offer several potential advantages over other temperature sensors. Measurements may be made near the melting point of the sensor material, potentially allowing monitoring of temperatures up to 3000 C. In addition, because no electrical insulation is required, shunting effects are avoided. Most attractive, however, is the ability to introduce acoustic discontinuities to the sensor, as this enables temperature measurements at several points along the sensor length. As discussed in this paper, the suite of temperature monitors offered by INL is not only available to ATR users, but also to users at other MTRs.

  17. Assessing Risk and Driving Risk Mitigation for First-of-a-Kind Advanced Reactors

    SciTech Connect (OSTI)

    John W. Collins

    2011-09-01T23:59:59.000Z

    Planning and decision making amidst programmatic and technological risks represent significant challenges for projects. This presentation addresses the four step risk-assessment process needed to determine clear path forward to mature needed technology and design, license, and construct advanced nuclear power plants, which have never been built before, including Small Modular Reactors. This four step process has been carefully applied to the Next Generation Nuclear Plant. STEP 1 - Risk Identification Risks are identified, collected, and categorized as technical risks, programmatic risks, and project risks, each of which result in cost and schedule impacts if realized. These include risks arising from the use of technologies not previously demonstrated in a relevant application. These risks include normal and accident scenarios which the SMR could experience including events that cause the disablement of engineered safety features (typically documented in Phenomena Identification Ranking Tables (PIRT) as produced with the Nuclear Regulatory Commission) and design needs which must be addressed to further detail the design. Product - Project Risk Register contained in a database with sorting, presentation, rollup, risk work off functionality similar to the NGNP Risk Management System . STEP 2 - Risk Quantification The risks contained in the risk register are then scored for probability of occurrence and severity of consequence, if realized. Here the scoring methodology is established and the basis for the scoring is well documented. Product - Quantified project risk register with documented basis for scoring. STEP 3 - Risk Handling Strategy Risks are mitigated by applying a systematic approach to maturing the technology through Research and Development, modeling, test, and design. A Technology Readiness Assessment is performed to determine baseline Technology Readiness Levels (TRL). Tasks needed to mature the technology are developed and documented in a roadmap. Product - Risk Handling Strategy. STEP 4 - Residual Risk Work off The risk handling strategy is entered into the Project Risk Allocation Tool (PRAT) to analyze each task for its ability to reduce risk. The result is risk-informed task prioritization. The risk handling strategy is captured in the Risk Management System, a relational database that provides conventional database utility, including data maintenance, archiving, configuration control, and query ability. The tool's Hierarchy Tree allows visualization and analyses of complex relationships between risks, risk mitigation tasks, design needs, and PIRTs. Product - Project Risk Allocation Tool and Risk Management System which depict project plan to reduce risk and current progress in doing so.

  18. INITIATORS AND TRIGGERING CONDITIONS FOR ADAPTIVE AUTOMATION IN ADVANCED SMALL MODULAR REACTORS

    SciTech Connect (OSTI)

    Katya L Le Blanc; Johanna h Oxstrand

    2014-04-01T23:59:59.000Z

    It is anticipated that Advanced Small Modular Reactors (AdvSMRs) will employ high degrees of automation. High levels of automation can enhance system performance, but often at the cost of reduced human performance. Automation can lead to human out-of the loop issues, unbalanced workload, complacency, and other problems if it is not designed properly. Researchers have proposed adaptive automation (defined as dynamic or flexible allocation of functions) as a way to get the benefits of higher levels of automation without the human performance costs. Adaptive automation has the potential to balance operator workload and enhance operator situation awareness by allocating functions to the operators in a way that is sensitive to overall workload and capabilities at the time of operation. However, there still a number of questions regarding how to effectively design adaptive automation to achieve that potential. One of those questions is related to how to initiate (or trigger) a shift in automation in order to provide maximal sensitivity to operator needs without introducing undesirable consequences (such as unpredictable mode changes). Several triggering mechanisms for shifts in adaptive automation have been proposed including: operator initiated, critical events, performance-based, physiological measurement, model-based, and hybrid methods. As part of a larger project to develop design guidance for human-automation collaboration in AdvSMRs, researchers at Idaho National Laboratory have investigated the effectiveness and applicability of each of these triggering mechanisms in the context of AdvSMR. Researchers reviewed the empirical literature on adaptive automation and assessed each triggering mechanism based on the human-system performance consequences of employing that mechanism. Researchers also assessed the practicality and feasibility of using the mechanism in the context of an AdvSMR control room. Results indicate that there are tradeoffs associated with each mechanism, but that some are more applicable to the AdvSMR domain. The two mechanisms that consistently improve performance in laboratory studies are operator initiated adaptive automation based on hierarchical task delegation and the Electroencephalogram(EEG) –based measure of engagement. Current EEG methods are intrusive and require intensive analysis; therefore it is not recommended for an AdvSMR control rooms at this time. Researchers also discuss limitations in the existing empirical literature and make recommendations for further research.

  19. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-05-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

  20. actinide burner reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    in this work, fission cross-sections on 233U, the main fissile isotope of the ThU fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on...

  1. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect (OSTI)

    Not Available

    1984-04-01T23:59:59.000Z

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  2. Stability analysis of the boiling water reactor : methods and advanced designs

    E-Print Network [OSTI]

    Hu, Rui, Ph. D. Massachusetts Institute of Technology

    2010-01-01T23:59:59.000Z

    Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

  3. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels 

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30T23:59:59.000Z

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, ...

  4. Application of advanced liquid metal reactors to the destruction of radioactive wastes

    SciTech Connect (OSTI)

    Karnesky, R.A.; Dobbin, K.D.; Jordheim, D.P.; Rawlins, J.A.; Wootan, D.W.

    1991-08-01T23:59:59.000Z

    The ability of a small fast reactor to destroy hazardous long lived minor actinide and fission product wastes is evaluated. It is determined that by using a novel technique wherein high energy neutrons leaking from the active core of the reactor are moderated by yttrium hydride located in the target assemblies, substantial amounts of long lived fission products can be destroyed and useful quantities of the beneficial isotope {sup 238}Pu can be produced by transmutation of the {sup 237}Np and {sub 241}Am minor actinide waste components. In addition it is shown that minor actinides recovered from spent Light Water Reactor fuel can be used to fuel such a reactor, increasing the amount of hazardous minor actinide and fission product wastes that can be destroyed. 9 refs., 4 figs., 2 tabs.

  5. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30T23:59:59.000Z

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, these technologies could...

  6. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

    SciTech Connect (OSTI)

    Guidez, J.; Martin, L. [Commissariat a l'Energie Atomique - CEA (France); Dupraz, R. [AREVA NP (France)

    2006-07-01T23:59:59.000Z

    The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

  7. Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards

    SciTech Connect (OSTI)

    Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01T23:59:59.000Z

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino detectors that were deployed. Finally, some preliminary results of our aboveground test will be shown. (authors)

  8. Radiation Resistance of XLPE Nano-dielectrics for Advanced Reactor Applications

    SciTech Connect (OSTI)

    Duckworth, Robert C [ORNL; Polyzos, Georgios [ORNL; Paranthaman, Mariappan Parans [ORNL; Aytug, Tolga [ORNL; Leonard, Keith J [ORNL; Sauers, Isidor [ORNL

    2014-01-01T23:59:59.000Z

    Recently there has been renewed interest in nuclear reactor safety, particularly as commercial reactors are approaching 40 years service and lifetime extensions are considered, as well as for new reactor building projects around the world. The materials that are currently used in cabling for instrumentation, reactor control, and communications include cross-linked polyethylene (XLPE), ethylene propylene rubber (EPR), polyvinyl chloride (PVC), neoprene, and chlorosulfonated polyethylene. While these materials show suitable radiation tolerance in laboratory tests, failures before their useful lifetime occur due to the combined environmental effects of radiation, temperature and moisture, or operation under abnormal conditions. In addition, the extended use of commercial reactors beyond their original service life places a greater demand on insulating materials to perform beyond their current ratings in these nuclear environments. Nanocomposite materials that are based on XLPE and other epoxy resins incorporating TiO2, MgO, SiO2, and Al2O3 nanoparticles are being fabricated using a novel in-situ method established at ORNL to demonstrate materials with increased resistance to radiation. As novel nanocomposite dielectric materials are developed, characterization of the non-irradiated and irradiated nanodielectrics will lead to a knowledge base that allow for dielectric materials to be engineered with specific nanoparticle additions for maximum benefit to wide-variety of radiation environments found in nuclear reactors. This paper presents the initial findings on the development of XLPE-based SiO2 nano-composite dielectrics in the context of electrical performance and radiation degradation.

  9. New Sensors for In-Pile Temperature Measurement at the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

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

    2011-09-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) in April 2007 to support U.S. research in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  10. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; J. E. Daw; K. G. Condie; S. Curtis Wilkins

    2009-09-01T23:59:59.000Z

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  11. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01T23:59:59.000Z

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  12. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    passive safety cooling systems. To develop an understandingthe passive safety cooling system and recommend an approachof Passive Safety Cooling Systems for Advanced Nuclear

  13. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    SciTech Connect (OSTI)

    Ruggles, A.E. [Oak Ridge National Lab., TN (United States)]|[Tennessee Univ., Knoxville, TN (United States); Cheng, L.Y. [Brookhaven National Lab., Upton, NY (United States); Dimenna, R.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Griffith, P. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Wilson, G.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1994-06-01T23:59:59.000Z

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  14. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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 significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

  15. AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor

    SciTech Connect (OSTI)

    T. Lambert; B. Grover; P. Guillermier; D. Moulinier; F. Imbault Huart

    2012-10-01T23:59:59.000Z

    AGR-2, the second irradiation of the US program for qualification of the NGNP fuel, is open to international participation within the scope of the Generation IV International Forum. In this frame, it includes in its multi-capsule irradiation rig an irradiation of French HTR fuel manufactured in the CAPRI line (GAIA facility at CEA/Cadarache and AREVA/CERCA compacting line at Romans). The AGR-2 irradiation is designed to place our first fabrications of HTR particles under operating conditions that are representative of ANTARES project while keeping close to the test range of the German fuel as much as possible, which is the reference in terms of irradiation behavior. A few batches of particles and 12 fuel compacts were produced and characterized in 2009 by CEA and CERCA. The fuel main characteristics are in conformity with our specifications and in compliance with INL requirements. The AGR-2 experiment is based on the design and devices used in the first experiment of the AGR program. The design makes it possible to monitor the irradiation conditions and in particular, the temperature, the power and the fission products released from fuel particles. The in pile equipment consists of a multi-capsule device designed to simultaneously irradiate six independent capsules with temperature control. The out-of-core part consists of the equipment for actively controlling temperature and measuring the fission products release on-line. The target conditions for the irradiation experiment were defined with the aim of comparing the results obtained under irradiation with German particles along with the objectives of reaching burn-up and fluence targets to validate the behavior of our fuel in a significant range (15% FIMA – 5 × 1025 n/m2 at 600 EFPD with centerline fuel temperature about 1100 degrees C). These conditions have to be representative of ANTARES project characteristics. These target conditions were compared with final results from neutron and thermal design studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.

  16. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01T23:59:59.000Z

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  17. SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS

    SciTech Connect (OSTI)

    Shamsuddin Ilias

    2005-02-03T23:59:59.000Z

    Inorganic membrane reactors offer the possibility of combining reaction and separation in a single operation at high temperatures to overcome the equilibrium limitations experienced in conventional reactor configurations. Such attractive features can be advantageously utilized in a number of potential commercial opportunities, which include dehydrogenation, hydrogenation, oxidative dehydrogenation, oxidation and catalytic decomposition reactions. However, to be cost effective, significant technological advances and improvements will be required to solve several key issues which include: (a) permselective thin solid film, (b) thermal, chemical and mechanical stability of the film at high temperatures, and (c) reactor engineering and module development in relation to the development of effective seals at high temperature and high pressure. In this project, we are working on the development and application of palladium and palladium-silver alloy thin-film composite membranes in membrane reactor-separator configuration for simultaneous production and separation of hydrogen and carbon dioxide at high temperature. From our research on Pd-composite membrane, we have demonstrated that the new membrane has significantly higher hydrogen flux with very high perm-selectivity than any of the membranes commercially available. The steam reforming of methane by equilibrium shift in Pd-composite membrane reactor is being studied to demonstrate the potential application of this new development. A two-dimensional, pseudo-homogeneous membrane-reactor model was developed to investigate the steam-methane reforming (SMR) reactions in a Pd-based membrane reactor. Radial diffusion was taken into consideration to account for the concentration gradient in the radial direction due to hydrogen permeation through the membrane. With appropriate reaction rate expressions, a set of partial differential equations was derived using the continuity equation for the reaction system. The equations were solved by finite difference method. The solution of the model equations is complicated by the coupled reactions. At the inlet, if there is no hydrogen, rate expressions become singular. To overcome this problem, the first element of the reactor was treated as a continuous stirred tank reactor (CSTR). Several alternative numerical schemes were implemented in the solution algorithm to get a converged, stable solution. The model was also capable of handling steam-methane reforming reactions under non-membrane condition and equilibrium reaction conversions. Some of the numerical results were presented in the previous report. To test the membrane reactor model, we fabricated Pd-stainless steel membranes in tubular configuration using electroless plating method coupled with osmotic pressure. Scanning Electron Microscopy (SEM) and Energy Dispersive X-ray (EDX) were used to characterize the fabricated Pd-film composite membranes. Gas-permeation tests were performed to measure the permeability of hydrogen, nitrogen and helium using pure gas. The membranes showed excellent perm-selectivity for hydrogen. This makes the Pd-composite membrane attractive for selective separation and recovery of H{sub 2} from mixed gases at elevated temperature.

  18. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  19. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    SciTech Connect (OSTI)

    Gehin, J.C.; Worley, B.A.; Renier, J.P. [Oak Ridge National Lab., TN (United States); Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01T23:59:59.000Z

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  20. The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

    SciTech Connect (OSTI)

    Cousseau, P.; Engelstad, R.; Henderson, D.L. [and others

    1997-10-01T23:59:59.000Z

    Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing.

  1. Advanced neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    E-Print Network [OSTI]

    Jang, Si Young

    2005-11-01T23:59:59.000Z

    was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). By increasing the thickness of the lead-bismuth alloy, the neutron spectra were shifted into lower energies by the scattering interactions of fast...

  2. advanced test reactor critical facility: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    test reactor critical facility First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Powerline Conductor...

  3. Programme A. Nuclear Power Subprogramme A.4 Technology Development for Advanced Reactor Lines

    E-Print Network [OSTI]

    De Cindio, Fiorella

    output, which improves hydrogen production efficiency and their free waste heat that can be used Title: Heat transfer behaviour and thermohydraulic code testing for supercritical water cooled reactors data for heat transfer; pressure drop, blowdown, natural circulation and stability for conditions

  4. Advances Towards Readily Deployable Antineutrino Detectors for Reactor Monitoring and Safeguards

    SciTech Connect (OSTI)

    Bowden, N S; Bernstein, A; Dazeley, S; Lund, J; Reyna, D; Sadler, L; Svoboda, R

    2008-06-05T23:59:59.000Z

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our LLNL/SNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale detector at tens of meters standoff from a commercial PWR. With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. Recently, we have investigated several technology paths that could allow such devices to be more readily deployed in the field. In particular, we have developed and fielded two new detectors; a low cost, non- flammable water based design; and a robust solid-state design based upon plastic scintillator. Here we will describe the tradeoffs inherent in these designs, and present results from their field deployments.

  5. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect (OSTI)

    none,

    1983-10-01T23:59:59.000Z

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  6. Low-Emissions Burner Technology using Biomass-Derived Liquid...

    Broader source: Energy.gov (indexed) [DOE]

    This factsheet describes a project that developed fuel-flexible, low-emissions burner technology capable of using biomass-derived liquid fuels, such as glycerin or fatty acids, as...

  7. Burner Designs and Controls for Variable Air Preheat Systems

    E-Print Network [OSTI]

    Lied, C. R.

    1981-01-01T23:59:59.000Z

    This paper will deal with various ways of reducing fuel costs for direct fired furnaces. Burner design relating to existing furnaces, new furnaces designed to operate initially on cold air with the ability to add preheated air in the future...

  8. J.R. Simplot: Burner Upgrade Project Improves Performance and...

    Broader source: Energy.gov (indexed) [DOE]

    how the J.R. Simplot Company saved energy and money by increasing the efficiency of the steam system in its potato processing plant in Caldwell, Idaho. J.R. Simplot: Burner Upgrade...

  9. 10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement

    SciTech Connect (OSTI)

    David E. Korns

    2012-05-01T23:59:59.000Z

    The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

  10. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  11. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    H. G. MacPherson The molten salt adventure Nuclear Scienceand P.F. Peterson, Molten-Salt-Cooled Advanced High-Clarno Assessment of candidate molten salt coolants for the

  12. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    Average Reynolds shellside crossflow 624 °C 90.6 (W/ m 2 °C)transfer from tubes in crossflow. Advances in Heat Transfercalc-init- *-------pressure drop crossflow bundle segment

  13. Advanced Safeguards Approaches for New Reprocessing Facilities

    SciTech Connect (OSTI)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Richard; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-06-24T23:59:59.000Z

    U.S. efforts to promote the international expansion of nuclear energy through the Global Nuclear Energy Partnership (GNEP) will result in a dramatic expansion of nuclear fuel cycle facilities in the United States. New demonstration facilities, such as the Advanced Fuel Cycle Facility (AFCF), the Advanced Burner Reactor (ABR), and the Consolidated Fuel Treatment Center (CFTC) will use advanced nuclear and chemical process technologies that must incorporate increased proliferation resistance to enhance nuclear safeguards. The ASA-100 Project, “Advanced Safeguards Approaches for New Nuclear Fuel Cycle Facilities,” commissioned by the NA-243 Office of NNSA, has been tasked with reviewing and developing advanced safeguards approaches for these demonstration facilities. Because one goal of GNEP is developing and sharing proliferation-resistant nuclear technology and services with partner nations, the safeguards approaches considered are consistent with international safeguards as currently implemented by the International Atomic Energy Agency (IAEA). This first report reviews possible safeguards approaches for the new fuel reprocessing processes to be deployed at the AFCF and CFTC facilities. Similar analyses addressing the ABR and transuranic (TRU) fuel fabrication lines at AFCF and CFTC will be presented in subsequent reports.

  14. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect (OSTI)

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

    2004-10-01T23:59:59.000Z

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

  15. Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect (OSTI)

    Lisa Harvego; Brion Bennett

    2011-11-01T23:59:59.000Z

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  16. Advanced light water reactor plants system 80+{trademark} design certification program. Annual progress report, October 1, 1993--September 30, 1994

    SciTech Connect (OSTI)

    Not Available

    1995-01-01T23:59:59.000Z

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80{sup +}{trademark} during the U.S. government`s 1994 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems. Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units and the System 80+ design form the basis of the Korean standardization program. The Nuclear Island portion of the System 80+ standard design has also been offered to the Republic of China, in response to their bid specification for an ALWR. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was docketed by the Nuclear Regulatory Commission (NRC) in May 1991 and a Draft Safety Evaluation Report (DSER) was issued in October 1992.

  17. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes Â… September 9, 2010

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r kiVP-^"^^?CONCEPTUALor.

  18. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes Â… September 9, 2010

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r kiVP-^"^^?CONCEPTUALor. 1 CASL

  19. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes Â… September 9, 2010

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r kiVP-^"^^?CONCEPTUALor. 1

  20. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes Â… September 9, 2010

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r kiVP-^"^^?CONCEPTUALor. 1Meeting

  1. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes Â… September 9, 2010

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r kiVP-^"^^?CONCEPTUALor.

  2. Advanced Fuel Performance: Modeling and Simulation Light Water Reactor Fuel Performance:

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041cloth DocumentationProducts (VAP) VAP7-0973 1BP-14 Power andAdvanced ComponentsenzymeAdvanced Fossil

  3. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  4. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    SciTech Connect (OSTI)

    Yoder, G.L.

    2005-10-03T23:59:59.000Z

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  5. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  6. Enhanced Combustion Low NOx Pulverized Coal Burner

    SciTech Connect (OSTI)

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

    2007-06-30T23:59:59.000Z

    For more than two decades, Alstom Power Inc. (Alstom) has developed a range of low cost, infurnace technologies for NOx emissions control for the domestic U.S. pulverized coal fired boiler market. This includes Alstom's internally developed TFS 2000{trademark} firing system, and various enhancements to it developed in concert with the U.S. Department of Energy. As of the date of this report, more than 270 units representing approximately 80,000 MWe of domestic coal fired capacity have been retrofit with Alstom low NOx technology. Best of class emissions range from 0.18 lb/MMBtu for bituminous coal to 0.10 lb/MMBtu for subbituminous coal, with typical levels at 0.24 lb/MMBtu and 0.13 lb/MMBtu, respectively. Despite these gains, NOx emissions limits in the U.S. continue to ratchet down for new and existing boiler equipment. On March 10, 2005, the Environmental Protection Agency (EPA) announced the Clean Air Interstate Rule (CAIR). CAIR requires 25 Eastern states to reduce NOx emissions from the power generation sector by 1.7 million tons in 2009 and 2.0 million tons by 2015. Low cost solutions to meet such regulations, and in particular those that can avoid the need for a costly selective catalytic reduction system (SCR), provide a strong incentive to continue to improve low NOx firing system technology to meet current and anticipated NOx control regulations. The overall objective of the work is to develop an enhanced combustion, low NOx pulverized coal burner, which, when integrated with Alstom's state-of-the-art, globally air staged low NOx firing systems will provide a means to achieve: Less than 0.15 lb/MMBtu NOx emissions when firing a high volatile Eastern or Western bituminous coal, Less than 0.10 lb/MMBtu NOx emissions when firing a subbituminous coal, NOx reduction costs at least 25% lower than the costs of an SCR, Validation of the NOx control technology developed through large (15 MWt) pilot scale demonstration, and Documentation required for economic evaluation and commercial application. During the project performance period, Alstom performed computational fluid dynamics (CFD) modeling and large pilot scale combustion testing in its Industrial Scale Burner Facility (ISBF) at its U.S. Power Plant Laboratories facility in Windsor, Connecticut in support of these objectives. The NOx reduction approach was to optimize near-field combustion to ensure that minimum NOx emissions are achieved with minimal impact on unburned carbon in ash, slagging and fouling, corrosion, and flame stability/turn-down. Several iterations of CFD and combustion testing on a Midwest coal led to an optimized design, which was extensively combustion tested on a range of coals. The data from these tests were then used to validate system costs and benefits versus SCR. Three coals were evaluated during the bench-scale and large pilot-scale testing tasks. The three coals ranged from a very reactive subbituminous coal to a moderately reactive Western bituminous coal to a much less reactive Midwest bituminous coal. Bench-scale testing was comprised of standard ASTM properties evaluation, plus more detailed characterization of fuel properties through drop tube furnace testing and thermogravimetric analysis. Bench-scale characterization of the three test coals showed that both NOx emissions and combustion performance are a strong function of coal properties. The more reactive coals evolved more of their fuel bound nitrogen in the substoichiometric main burner zone than less reactive coal, resulting in the potential for lower NOx emissions. From a combustion point of view, the more reactive coals also showed lower carbon in ash and CO values than the less reactive coal at any given main burner zone stoichiometry. According to bench-scale results, the subbituminous coal was found to be the most amenable to both low NOx, and acceptably low combustibles in the flue gas, in an air staged low NOx system. The Midwest bituminous coal, by contrast, was predicted to be the most challenging of the three coals, with the Western bituminous coal predicted to beh

  7. Dual-water mixture fuel burner

    DOE Patents [OSTI]

    Brown, Thomas D. (Finleyville, PA); Reehl, Douglas P. (Pittsburgh, PA); Walbert, Gary F. (Library, PA)

    1986-08-05T23:59:59.000Z

    A coal-water mixture (CWM) burner includes a conically shaped rotating cup into which fuel comprised of coal particles suspended in a slurry is introduced via a first, elongated inner tube coupled to a narrow first end portion of the cup. A second, elongated outer tube is coaxially positioned about the first tube and delivers steam to the narrow first end of the cup. The fuel delivery end of the inner first tube is provided with a helical slot on its lateral surface for directing the CWM onto the inner surface of the rotating cup in the form of a uniform, thin sheet which, under the influence of the cup's centrifugal force, flows toward a second, open, expanded end portion of the rotating cup positioned immediately adjacent to a combustion chamber. The steam delivered to the rotating cup wets its inner surface and inhibits the coal within the CWM from adhering to the rotating cup. A primary air source directs a high velocity air flow coaxially about the expanded discharge end of the rotating cup for applying a shear force to the CWM in atomizing the fuel mixture for improved combustion. A secondary air source directs secondary air into the combustion chamber adjacent to the outlet of the rotating cup at a desired pitch angle relative to the fuel mixture/steam flow to promote recirculation of hot combustion gases within the ignition zone for increased flame stability.

  8. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

    2009-05-01T23:59:59.000Z

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in the graphite fuel compacts versus EFPD, the P/T ratio was calculated to be 5.3, which is unacceptable given the fuel compact temperature control requirement. To flatten the FPD profile versus EFPDs, two proposed options are – (a) add fertile (232Th) particles to the fuel compact and (b) add burnable absorber (B4C) to the graphite holder. The effectiveness of these two proposed options to flatten the FPD profile versus EFPDs were investigated and the results are compared in this study.

  9. Modeling prismatic HTGRs with U.S. N.R.C advanced gas reactor evaluator (AGREE)

    SciTech Connect (OSTI)

    Seker, V.; Drzewiecki, T.; Downar, T. [Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington, DC (United States)

    2012-07-01T23:59:59.000Z

    A core fluids and heat transfer model has been developed for the prismatic high temperature gas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The core fluids modeling relies on a subchannel approach in which the primary coolant flow path through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation for the core fluids model has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. The core level heat transfer model is based on a triangular finite volume method, where the base triangle is one sixth of the prismatic block. In order to improve the spatial accuracy at this level, a triangular refinement method was also implemented. The fuel compact temperature is calculated by a cylindrical conduction model which is implicitly coupled to the triangular core level model. The preliminary verification of the model was performed by comparing AGREE to a finite element code COMSOL by analyzing the MHTGR core heat transfer. Further verification and validation is currently an ongoing effort. (authors)

  10. Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

    SciTech Connect (OSTI)

    B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

    2013-09-01T23:59:59.000Z

    This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

  11. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    SciTech Connect (OSTI)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29T23:59:59.000Z

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  12. Technical Readiness and Gaps Analysis of Commercial Optical Materials and Measurement Systems for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Anheier, Norman C.; Suter, Jonathan D.; Qiao, Hong (Amy); Andersen, Eric S.; Berglin, Eric J.; Bliss, Mary; Cannon, Bret D.; Devanathan, Ramaswami; Mendoza, Albert; Sheen, David M.

    2013-08-06T23:59:59.000Z

    This report intends to support Department of Energy’s Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap and industry stakeholders by evaluating optical-based instrumentation and control (I&C) concepts for advanced small modular reactor (AdvSMR) applications. These advanced designs will require innovative thinking in terms of engineering approaches, materials integration, and I&C concepts to realize their eventual viability and deployability. The primary goals of this report include: 1. Establish preliminary I&C needs, performance requirements, and possible gaps for AdvSMR designs based on best available published design data. 2. Document commercial off-the-shelf (COTS) optical sensors, components, and materials in terms of their technical readiness to support essential AdvSMR in-vessel I&C systems. 3. Identify technology gaps by comparing the in-vessel monitoring requirements and environmental constraints to COTS optical sensor and materials performance specifications. 4. Outline a future research, development, and demonstration (RD&D) program plan that addresses these gaps and develops optical-based I&C systems that enhance the viability of future AdvSMR designs. The development of clean, affordable, safe, and proliferation-resistant nuclear power is a key goal that is documented in the Nuclear Energy Research and Development Roadmap. This roadmap outlines RD&D activities intended to overcome technical, economic, and other barriers, which currently limit advances in nuclear energy. These activities will ensure that nuclear energy remains a viable component to this nation’s energy security.

  13. Flame quality monitor system for fixed firing rate oil burners

    DOE Patents [OSTI]

    Butcher, Thomas A. (Pt. Jefferson, NY); Cerniglia, Philip (Moriches, NY)

    1992-01-01T23:59:59.000Z

    A method and apparatus for determining and indicating the flame quality, or efficiency of the air-fuel ratio, in a fixed firing rate heating unit, such as an oil burning furnace, is provided. When the flame brightness falls outside a preset range, the flame quality, or excess air, has changed to the point that the unit should be serviced. The flame quality indicator output is in the form of lights mounted on the front of the unit. A green light indicates that the flame is about in the same condition as when the burner was last serviced. A red light indicates a flame which is either too rich or too lean, and that servicing of the burner is required. At the end of each firing cycle, the flame quality indicator goes into a hold mode which is in effect during the period that the burner remains off. A yellow or amber light indicates that the burner is in the hold mode. In this mode, the flame quality lights indicate the flame condition immediately before the burner turned off. Thus the unit can be viewed when it is off, and the flame condition at the end of the previous firing cycle can be observed.

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

    SciTech Connect (OSTI)

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

    2001-09-04T23:59:59.000Z

    It is well understood that the stability of axial diffusion flames is dependent on the mixing behavior of the fuel and combustion air streams. Combustion aerodynamic texts typically describe flame stability and transitions from laminar diffusion flames to fully developed turbulent flames as a function of increasing jet velocity. Turbulent diffusion flame stability is greatly influenced by recirculation eddies that transport hot combustion gases back to the burner nozzle. This recirculation enhances mixing and heats the incoming gas streams. Models describing these recirculation eddies utilize conservation of momentum and mass assumptions. Increasing the mass flow rate of either fuel or combustion air increases both the jet velocity and momentum for a fixed burner configuration. Thus, differentiating between gas velocity and momentum is important when evaluating flame stability under various operating conditions. The research efforts described herein are part of an ongoing project directed at evaluating the effect of flame aerodynamics on NO{sub x} emissions from coal fired burners in a systematic manner. This research includes both experimental and modeling efforts being performed at the University of Arizona in collaboration with Purdue University. The objective of this effort is to develop rational design tools for optimizing low NO{sub x} burners. Experimental studies include both cold-and hot-flow evaluations of the following parameters: primary and secondary inlet air velocity, coal concentration in the primary air, coal particle size distribution and flame holder geometry. Hot-flow experiments will also evaluate the effect of wall temperature on burner performance.

  15. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10T23:59:59.000Z

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  16. A Framework to Expand and Advance Probabilistic Risk Assessment to Support Small Modular Reactors

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Robert Nourgaliev; Cherie Phelan; Diego Mandelli; Kellie Kvarfordt; Robert Youngblood

    2012-09-01T23:59:59.000Z

    During the early development of nuclear power plants, researchers and engineers focused on many aspects of plant operation, two of which were getting the newly-found technology to work and minimizing the likelihood of perceived accidents through redundancy and diversity. As time, and our experience, has progressed, the realization of plant operational risk/reliability has entered into the design, operation, and regulation of these plants. But, to date, we have only dabbled at the surface of risk and reliability technologies. For the next generation of small modular reactors (SMRs), it is imperative that these technologies evolve into an accepted, encompassing, validated, and integral part of the plant in order to reduce costs and to demonstrate safe operation. Further, while it is presumed that safety margins are substantial for proposed SMR designs, the depiction and demonstration of these margins needs to be better understood in order to optimize the licensing process.

  17. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOE Patents [OSTI]

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05T23:59:59.000Z

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  18. High Efficiency Burners by Retrofit - A Simple Inexpensive Way to Improve Combustion Efficiency

    E-Print Network [OSTI]

    Rogers, W. T.

    1980-01-01T23:59:59.000Z

    Existing direct fired process heaters and steam boilers can have their efficiencies remarkably improved, and thus cut the fuel bill, by conversion from conventional type natural draft burners to high intensity, "forced draft" type burners...

  19. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    SciTech Connect (OSTI)

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01T23:59:59.000Z

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  20. New Sensors for the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson; Keith G. Condie; Joshua E. Daw; Heng Ban; Brandon Fox; Gordon Kohse

    2009-06-01T23:59:59.000Z

    A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the selection strategy of what instrumentation is needed, and the program generated for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available to users of the ATR NSUF with data from irradiation tests using these sensors. In addition, progress is reported on current research efforts to provide users advanced methods for detecting temperature, fuel thermal conductivity, and changes in sample geometry.

  1. Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the YouTube platform is alwaysISO 50001EnergyNewsletter AdvancedWindow andCivilEnergy

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

    SciTech Connect (OSTI)

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

    2001-06-01T23:59:59.000Z

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

  3. Slurry burner for mixture of carbonaceous material and water

    DOE Patents [OSTI]

    Nodd, D.G.; Walker, R.J.

    1985-11-05T23:59:59.000Z

    The present invention is intended to overcome the limitations of the prior art by providing a fuel burner particularly adapted for the combustion of carbonaceous material-water slurries which includes a stationary high pressure tip-emulsion atomizer which directs a uniform fuel into a shearing air flow as the carbonaceous material-water slurry is directed into a combustion chamber, inhibits the collection of unburned fuel upon and within the atomizer, reduces the slurry to a collection of fine particles upon discharge into the combustion chamber, and regulates the operating temperature of the burner as well as primary air flow about the burner and into the combustion chamber for improved combustion efficiency, no atomizer plugging and enhanced flame stability.

  4. Description of TASHA: Thermal Analysis of Steady-State-Heat Transfer for the Advanced Neutron Source Reactor

    SciTech Connect (OSTI)

    Morris, D.G.; Chen, N.C.; Nelson, W.R.; Yoder, G.L.

    1996-10-01T23:59:59.000Z

    This document describes the code used to perform Thermal Analysis of Steady-State-Heat-Transfer for the Advanced Neutron Source (ANS) Reactor (TASHA). More specifically, the code is designed for thermal analysis of the fuel elements. The new code reflects changes to the High Flux Isotope Reactor steady-state thermal-hydraulics code. These changes were aimed at both improving the code`s predictive ability and allowing statistical thermal-hydraulic uncertainty analysis to be performed. A significant portion of the changes were aimed at improving the correlation package in the code. This involved incorporating more recent correlations for both single-phase flow and two-phase flow thermal limits, including the addition of correlations to predict the phenomenon of flow excursion. Since the code was to be used in the design of the ANS, changes were made to allow the code to predict limiting powers for a variety of thermal limits, including critical heat flux, flow excursion, incipient boiling, oxide spallation, maximum centerline temperature, and surface temperature equal to the saturation temperature. Statistical uncertainty analysis also required several changes to the code itself as well as changes to the code input format. This report describes these changes in enough detail to allow the reader to interpret code results and also to understand where the changes were made in the code programming. This report is not intended to be a stand alone report for running the code, however, and should be used in concert with the two previous reports published on the original code. Sample input and output files are also included to help accomplish these goals. In addition, a section is included that describes requirements for a new, more modem code that the project planned to develop.

  5. Development of quick repairing technique for ceramic burner in hot stove of blast furnace

    SciTech Connect (OSTI)

    Kondo, Atsushi; Doura, Kouji; Nakamura, Hirofumi [Sumitomo Metal Industries, Ltd., Wakayama (Japan). Wakayama Steel Works

    1997-12-31T23:59:59.000Z

    Refractories of ceramic burner in hot stoves at Wakayama No. 4 blast furnace were damaged. There are only three hot stoves, so repairing must be done in a short. Therefore, a quick repairing technique for ceramic burners has been developed, and two ceramic burners were repaired in just 48 hours.

  6. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    SciTech Connect (OSTI)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01T23:59:59.000Z

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

  7. An Advanced Computational Scheme for the Optimization of 2D Radial Reflectors in Pressurized Water Reactors

    E-Print Network [OSTI]

    Thomas Clerc; Alain Hébert; Hadrien Leroyer; Jean-Philippe Argaud; Bertrand Bouriquet; Agélique Ponçot

    2014-05-12T23:59:59.000Z

    This paper presents a computational scheme for the determination of equivalent 2D multi-group heterogeneous reflectors in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de donn\\'{e}es et Aide \\`{a} l'Optimisation (ADAO)" of the SALOME platform developed at \\'{E}lectricit\\'{e} De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first validation of the computational scheme is made using the OPTEX reflector model developed at \\'{E}cole Polytechnique de Montr\\'{e}al (EPM). As a result, we obtain 2D multi-group, spatially heterogeneous 2D reflectors, using both diffusion or $\\text{SP}_{\\text{N}}$ operators. We observe important improvements of the power discrepancies distribution over the core when using reflectors computed with the proposed computational scheme, and the $\\text{SP}_{\\text{N}}$ operator enables additional improvements.

  8. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect (OSTI)

    Ware, A.G.

    1994-07-01T23:59:59.000Z

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  9. An Advanced Computational Scheme for the Optimization of 2D Radial Reflectors in Pressurized Water Reactors

    E-Print Network [OSTI]

    Clerc, Thomas; Leroyer, Hadrien; Argaud, Jean-Philippe; Bouriquet, Bertrand; Ponçot, Agélique

    2014-01-01T23:59:59.000Z

    This paper presents a computational scheme for the determination of equivalent 2D multi-group heterogeneous reflectors in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de donn\\'{e}es et Aide \\`{a} l'Optimisation (ADAO)" of the SALOME platform developed at \\'{E}lectricit\\'{e} De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first validation of the computational scheme is made using the OPTEX reflector model developed at \\'{E}cole Polytechnique de Montr\\'{e}al (EPM). As a result, we obtain 2D multi-group, spatially heterogeneous 2D reflectors, using both diffusion or $\\text{SP}_{\\text{N}}$ operators. We observe important improvements of the power discrepancies distribution over the cor...

  10. Some Specific CASL Requirements for Advanced Multiphase Flow Simulation of Light Water Reactors

    SciTech Connect (OSTI)

    R. A. Berry

    2010-11-01T23:59:59.000Z

    Because of the diversity of physical phenomena occuring in boiling, flashing, and bubble collapse, and of the length and time scales of LWR systems, it is imperative that the models have the following features: • Both vapor and liquid phases (and noncondensible phases, if present) must be treated as compressible. • Models must be mathematically and numerically well-posed. • The models methodology must be multi-scale. A fundamental derivation of the multiphase governing equation system, that should be used as a basis for advanced multiphase modeling in LWR coolant systems, is given in the Appendix using the ensemble averaging method. The remainder of this work focuses specifically on the compressible, well-posed, and multi-scale requirements of advanced simulation methods for these LWR coolant systems, because without these are the most fundamental aspects, without which widespread advancement cannot be claimed. Because of the expense of developing multiple special-purpose codes and the inherent inability to couple information from the multiple, separate length- and time-scales, efforts within CASL should be focused toward development of a multi-scale approaches to solve those multiphase flow problems relevant to LWR design and safety analysis. Efforts should be aimed at developing well-designed unified physical/mathematical and high-resolution numerical models for compressible, all-speed multiphase flows spanning: (1) Well-posed general mixture level (true multiphase) models for fast transient situations and safety analysis, (2) DNS (Direct Numerical Simulation)-like models to resolve interface level phenmena like flashing and boiling flows, and critical heat flux determination (necessarily including conjugate heat transfer), and (3) Multi-scale methods to resolve both (1) and (2) automatically, depending upon specified mesh resolution, and to couple different flow models (single-phase, multiphase with several velocities and pressures, multiphase with single velocity and pressure, etc.) A unified, multi-scale approach is advocated to extend the necessary foundations and build the capability to simultaneously solve the fluid dynamic interface problems (interface resolution) as well as multiphase mixtures (homogenization).

  11. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect (OSTI)

    Morreale, A.C.; Luxat, J.C. [McMaster University, 1280 Main St. W. Hamilton, Ontario, L8S 4L7 (Canada); Friedlander, Y. [AMEC-NSS Ltd., 700 University Ave. 4th Floor, Toronto, Ontario, M5G 1X6 (Canada)

    2013-07-01T23:59:59.000Z

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  12. 2012 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2011 through October 31, 2012. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance issues Discussion of the facility’s environmental impacts During the 2012 permit year, approximately 183 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  13. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect (OSTI)

    NONE

    1997-05-01T23:59:59.000Z

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  14. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2012-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  15. DOE/NE robotics for advanced reactors. Bimonthly progress report, October--November 1991

    SciTech Connect (OSTI)

    Not Available

    1991-12-31T23:59:59.000Z

    This document details activities during this reporting period. The Michigan group has developed, built, and tested a general purpose interface circuit for DC motors and encoders. This interface is based on an advanced microchip, the HCTL 1100 manufactured by Hewlett Packard. The HCTL 1100 can be programmed by a host computer in real-time, allowing sophisticated motion control for DC motors. At the University of Florida, work on modeling the details of the seismic isolators and the jack mechanism has been completed. A separate 3D solid view of the seismic isolator floor, with the full set of isolators shown in detail, has been constructed within IGRIP. ORNL led the robotics team at the ALMR review meeting. Discussions were held with General Electric (GE) engineers and contractors on the robotic needs for the ALMR program. The Tennessee group has completed geometric modeling of the Andros Mark VI mobile platform with two fixed tracks and for articulated tracks, the give degree-of-freedom manipulator and its end-effector, and two cameras. A graphical control of panel was developed which allow the user to operate the simulated robot. The University of Texas team visited ORNL to complete the implementation of computed-torque controller on the CESARm manipulator. This controller was previously developed and computer simulations were carried out specifically for the CESARm robot.

  16. Study of the Effects of Ambient Conditions Upon the Performance of Fan Powered, Infrared Natural Gas Burners

    SciTech Connect (OSTI)

    Clark Atlanta University

    2002-12-02T23:59:59.000Z

    The objective of this investigation was to characterize the operation of a fan-powered, infrared burner (IR burner) at various gas compositions and ambient conditions, develop numerical model to simulate the burner performances, and provide design guidelines for appliances containing PIR burners for satisfactory performance.

  17. Fire suppression efficiency screening using a counterflow cylindrical burner

    SciTech Connect (OSTI)

    Yang, J.C.; Donnelly, M.K.; Prive, N.; Grosshandler, W.L.

    1999-07-01T23:59:59.000Z

    The design and validation of a counterflow cylindrical burner for fire suppression efficiency screening are described. The stability limits of the burner were mapped using various fuel (propane) and oxidizer (air) flows. The stability envelopes compared favorably with those reported in the literature. The apparatus was characterized using inert gases (argon, helium, and nitrogen), and the relative fire suppression efficiency ranking of these three gases was found to be commensurate with that from cup-burner tests. For liquid suppression experiments, a piezoelectric droplet generator was used to form droplets (<100 {micro}m). Water was used as a representative liquid suppressant to study the feasibility of using such a burner for screening liquid agents. Extinction was facilitated with the addition of water droplets, and the effect of water became more pronounced when its application rate was increased. Suppression experiments using water with and without nitrogen dilution in the oxidizer stream were also performed. Flame extinction due to the combined effect of water and nitrogen dilution was demonstrated.

  18. BURNER DEVELOPMENT AND OPERABILITY ISSUES ASSOCIATED WITH STEADY FLOWING SYNGAS

    E-Print Network [OSTI]

    Lieuwen, Timothy C.

    BURNER DEVELOPMENT AND OPERABILITY ISSUES ASSOCIATED WITH STEADY FLOWING SYNGAS FIRED COMBUSTORS-Mu¨nchen, Garching, Germany This article addresses the impact of syngas fuel composition on combustor blowout, flash flashback mechanisms are present in swirling flows, and the key thermophysical properties of a syngas

  19. INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHENOMENA IN ADVANCED GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHE

    2006-09-01T23:59:59.000Z

    INL LDRD funded research was conducted at MIT to experimentally characterize mixed convection heat transfer in gas-cooled fast reactor (GFR) core channels in collaboration with INL personnel. The GFR for Generation IV has generated considerable interest and is under development in the U.S., France, and Japan. One of the key candidates is a block-core configuration first proposed by MIT, has the potential to operate in Deteriorated Turbulent Heat Transfer (DTHT) regime or in the transition between the DTHT and normal forced or laminar convection regime during post-loss-of-coolant accident (LOCA) conditions. This is contrary to most industrial applications where operation is in a well-defined and well-known turbulent forced convection regime. As a result, important new need emerged to develop heat transfer correlations that make possible rigorous and accurate predictions of Decay Heat Removal (DHR) during post LOCA in these regimes. Extensive literature review on these regimes was performed and a number of the available correlations was collected in: (1) forced laminar, (2) forced turbulent, (3) mixed convection laminar, (4) buoyancy driven DTHT and (5) acceleration driven DTHT regimes. Preliminary analysis on the GFR DHR system was performed and using the literature review results and GFR conditions. It confirmed that the GFR block type core has a potential to operate in the DTHT regime. Further, a newly proposed approach proved that gas, liquid and super critical fluids all behave differently in single channel under DTHT regime conditions, thus making it questionable to extrapolate liquid or supercritical fluid data to gas flow heat transfer. Experimental data were collected with three different gases (nitrogen, helium and carbon dioxide) in various heat transfer regimes. Each gas unveiled different physical phenomena. All data basically covered the forced turbulent heat transfer regime, nitrogen data covered the acceleration driven DTHT and buoyancy driven DTHT, helium data covered the mixed convection laminar, acceleration driven DTHT and the laminar to turbulent transition regimes and carbon dioxide data covered the returbulizing buoyancy driven DTHT and non-returbulizing buoyancy induced DTHT. The validity of the data was established using the heat balance and the uncertainty analysis. Based on experimental data, the traditional threshold for the DTHT regime was updated to account for phenomena observed in the facility and a new heat transfer regime map was proposed. Overall, it can be stated that substantial reduction of heat transfer coefficient was observed in DTHT regime, which will have significant impact on the core and DHR design of passive GFR. The data were compared to the large number of existing correlations. None of the mixed convection laminar correlation agreed with the data. The forced turbulent and the DTHT regime, Celeta et al. correlation showed the best fit with the data. However, due to larger ratio of the MIT facility compared to the Celeta et al. facility and the returbuliziation due to the gas characteristics, the correlation sometimes under-predicts the heat transfer coefficient. Also, since Celeta et al. correlation requires the information of the wall temperature to evaluate the heat transfer coefficient, it is difficult to apply this correlation directly for predicting the wall temperature. Three new sets of correlation that cover all heat transfer regimes were developed. The bas

  20. Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors

    SciTech Connect (OSTI)

    Unal, Cetin [Los Alamos National Laboratory; Williams, Brian [Los Alamos National Laboratory; Mc Clure, Patrick [Los Alamos National Laboratory; Nelson, Ralph A [IDAHO NATIONAL LAB

    2010-01-01T23:59:59.000Z

    Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M&S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for cost saving purposes by showing further testing wold not enhance the quality of the validation of predictive tools. The proposed methodology is at a conceptual level. When matured and if considered favorably by the stakeholders, it could serve as a new framework for the next generation of the best estimate plus uncertainty licensing methodology that USNRC developed previously. In order to come to that level of maturity it is necessary to communicate the methodology to scientific, design and regulatory stakeholders for discussion and debates. This paper is the first step to establish this communication.

  1. A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE

    SciTech Connect (OSTI)

    Jorge Navarro

    2013-12-01T23:59:59.000Z

    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 for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

  2. Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

    SciTech Connect (OSTI)

    Wachs, G. W.

    1998-09-01T23:59:59.000Z

    The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (MOX) fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (ATR) to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).

  3. Auxiliary reactor for a hydrocarbon reforming system

    DOE Patents [OSTI]

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17T23:59:59.000Z

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  4. DOE Seeks to Invest up to $15 Million in Funding for Nuclear...

    Energy Savers [EERE]

    following areas: Used Fuel Separations Technology, Advanced Nuclear Fuel Development, Fast Burner Reactors and Advanced Transmutation Systems, Advanced Fuel Cycle Systems...

  5. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01T23:59:59.000Z

    The United States Department of Energy, 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 decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  6. Advances

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041cloth DocumentationProducts (VAP) VAP7-0973 1BP-14Scripting for Advanced Workflows Jack

  7. Downhole burner systems and methods for heating subsurface formations

    DOE Patents [OSTI]

    Farmayan, Walter Farman (Houston, TX); Giles, Steven Paul (Damon, TX); Brignac, Jr., Joseph Phillip (Katy, TX); Munshi, Abdul Wahid (Houston, TX); Abbasi, Faraz (Sugarland, TX); Clomburg, Lloyd Anthony (Houston, TX); Anderson, Karl Gregory (Missouri City, TX); Tsai, Kuochen (Katy, TX); Siddoway, Mark Alan (Katy, TX)

    2011-05-31T23:59:59.000Z

    A gas burner assembly for heating a subsurface formation includes an oxidant conduit, a fuel conduit, and a plurality of oxidizers coupled to the oxidant conduit. At least one of the oxidizers includes a mix chamber for mixing fuel from the fuel conduit with oxidant from the oxidant conduit, an igniter, and a shield. The shield includes a plurality of openings in communication with the oxidant conduit. At least one flame stabilizer is coupled to the shield.

  8. Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs

    SciTech Connect (OSTI)

    Chang 'Apollo', Chen [Apollo Consulting, Inc., Surprise, AZ 85374-4605 (United States)

    2006-07-01T23:59:59.000Z

    For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to search for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)

  9. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect (OSTI)

    M. Chen; CM Regan; D. Noe

    2006-01-09T23:59:59.000Z

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  10. Pollutant Exposures from Natural Gas Cooking Burners: A Simulation-Based Assessment for Southern California

    E-Print Network [OSTI]

    Logue, Jennifer M.

    2014-01-01T23:59:59.000Z

    P. Sullivan (2009). Natural Gas Variability in California:Singer (2012). Impact of Natural Gas Appliances on PollutantPollutant Exposures in Natural Gas Cooking Burners, LBNL

  11. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01T23:59:59.000Z

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  12. TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis

    E-Print Network [OSTI]

    Griggs, D. P.

    1984-01-01T23:59:59.000Z

    The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

  13. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect (OSTI)

    Dautel, W.A.

    1996-10-01T23:59:59.000Z

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  14. CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced. C o w l i t z C o . C l a r k C o

  15. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    SciTech Connect (OSTI)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01T23:59:59.000Z

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

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

    E-Print Network [OSTI]

    Cvoro, Valentina

    Coal firing power stations represent the second largest source of global NOx emissions. The current practice of predicting likely exit NOx levels from multi-burner furnaces on the basis of single burner test rig data has been proven inadequate...

  17. Development, Application and Performance of Venturi Register L. E. A. Burner System for Firing Oil and Gas Fuels 

    E-Print Network [OSTI]

    Cawte, A. D.

    1979-01-01T23:59:59.000Z

    as CEA Combustion, Ltd., to develop a more efficient suspended - flame burner. Subsequently, the CEGB (Central Electric Generating Board) in Great Britain developed standards for register type burners installed in fossil fuel fired electric generating...

  18. U.S. Department Of Energy Advanced Small Modular Reactor R&D Program: Instrumentation, Controls, and Human-Machine Interface (ICHMI) Pathway

    SciTech Connect (OSTI)

    Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2013-01-01T23:59:59.000Z

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of modern ICHMI technology. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, several DOE programs have substantial ICHMI RD&D elements within their respective research portfolios. This paper describes current ICHMI research in support of advanced small modular reactors. The objectives that can be achieved through execution of the defined RD&D are to provide optimal technical solutions to critical ICHMI issues, resolve technology gaps arising from the unique measurement and control characteristics of advanced reactor concepts, provide demonstration of needed technologies and methodologies in the nuclear power application domain, mature emerging technologies to facilitate commercialization, and establish necessary technical evidence and application experience to enable timely and predictable licensing. 1 Introduction Instrumentation, controls, and human-machine interfaces are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface (ICHMI) systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of m

  19. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  20. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  1. Proceedings of the U. S. Nuclear Regulatory Commission review group conference on advanced instrumentation research for reactor safety held at Oak Ridge National Laboratory on July 29-31, 1980. Conference proceedings

    SciTech Connect (OSTI)

    Hon, A.L.; Basdekas, D.; Hsu, Y.Y.; Kondic, N.; Van Houten, R.

    1980-12-01T23:59:59.000Z

    The report compiles the technical presentations during the Advanced Instrumentation Research for the Reactor Safety Review Group Meeting held in July 1980. The three-day meeting covered the Power Plant Instrumentation, Two-Phase Flow Instrumentation, Fuel Behavior Research Instrumentation and Advanced Reactor Instrumentation research programs sponsored by the U.S. Nuclear Regulatory Commission, Division of Reactor Safety Research. In addition, two invited papers from the nuclear industry were also presented. The conference is held each year to review the up-to-date instrumentation research results by the contractors. It also provides the opportunity for the researchers and experts to exchange experience on advanced instrumentation development. The report serves as a vehicle to disseminate the state-of-the-art information to the research community and the nuclear industry.

  2. Slurry burner for mixture of carbonaceous material and water

    DOE Patents [OSTI]

    Nodd, Dennis G. (West Mifflin, PA); Walker, Richard J. (Bethel Park, PA)

    1987-01-01T23:59:59.000Z

    A carbonaceous material-water slurry burner includes a high pressure tip-emulsion atomizer for directing a carbonaceous material-water slurry into a combustion chamber for burning therein without requiring a support fuel or oxygen enrichment of the combustion air. Introduction of the carbonaceous material-water slurry under pressure forces it through a fixed atomizer wherein the slurry is reduced to small droplets by mixing with an atomizing air flow and directed into the combustion chamber. The atomizer includes a swirler located immediately adjacent to where the fuel slurry is introduced into the combustion chamber and which has a single center channel through which the carbonaceous material-water slurry flows into a plurality of diverging channels continuous with the center channel from which the slurry exits the swirler immediately adjacent to an aperture in the combustion chamber. The swirler includes a plurality of slots around its periphery extending the length thereof through which the atomizing air flows and by means of which the atomizing air is deflected so as to exert a maximum shear force upon the carbonaceous material-water slurry as it exits the swirler and enters the combustion chamber. A circulating coolant system or boiler feed water is provided around the periphery of the burner along the length thereof to regulate burner operating temperature, eliminate atomizer plugging, and inhibit the generation of sparklers, thus increasing combustion efficiency. A secondary air source directs heated air into the combustion chamber to promote recirculation of the hot combustion gases within the combustion chamber.

  3. CFCC radiant burner assessment. Final report, April 1, 1992--July 31, 1994

    SciTech Connect (OSTI)

    Schweizer, S.; Sullivan, J.

    1994-11-01T23:59:59.000Z

    The objective of this work was to identify methods of improving the performance of gas-fired radiant burners through the use of Continuous Fiber Ceramic Composites (CFCCs). Methods have been identified to improve the price and performance characteristics of the porous surface burner. Results are described.

  4. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect (OSTI)

    Not Available

    1994-01-15T23:59:59.000Z

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  5. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg; Sean R. Morrell

    2012-09-01T23:59:59.000Z

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

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

    SciTech Connect (OSTI)

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

    1998-10-01T23:59:59.000Z

    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.

  7. Check Burner Air to Fuel Ratios | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the YouTube platformBuilding RemovalCSSDepartment of Energy5-4-20129Burner Air to Fuel

  8. Development and testing of a high efficiency advanced coal combustor: Phase III, Industrial boiler retrofit. Quarterly technical progress report No. 14, January 1, 1995--March 31, 1995

    SciTech Connect (OSTI)

    Patel, R.L.; Borio, R. [ABB Combustion Engineering Systems, Windsor, CT (United States). U.S. Power Plant Labs.; Scaroni, A.W.; Miller, B.G. [Pennsylvania State Univ., PA (United States); McGowan, J.G. [Massachusetts Univ., MA (United States)

    1995-04-28T23:59:59.000Z

    The objective of this project is to retrofit the previously developed High Efficiency Advanced Coal Combustor (HEACC) to a standard gas/oil designed industrial boiler to assess the technical and economic viability of displacing premium fuels with microfine coal. This report documents the technical aspects of this project during the fourteenth quarter (January `95 through March `95) of the program. The ABB project team met with cognizant DOE-PETC and Penn State personnel on February 15, 1995 at Penn State to discuss our ideas for a new burner (RSFC-based) to replace the HEACC burner prior to the long term ({approximately}1000 hrs) demonstration phase of this project. The main reasons for the proposed new burner were to improve combustion efficiencies and NO{sub x} reduction. Recent, experience at MIT with 5 million Btu/hr coal firing experiments on RSFC burner have shown remarkable performance. Results indicate that RSFC-based burner has the potential to produce lower NO{sub x} and higher carbon conversion efficiencies than the HEACC burner. M.I.T. developed the RSFC burner and obtained a patent for the concept. A decision was made to go with the new, RSFC-based burner during 1000 hr demonstration. ABB-CE will fund the costs ({approximately}$50K) for design/fabrication of the proposed new burner. Penn State plans to improve coal handling by installation of a gravimetric feeder and redesign/installation of a mass flow bottom on the surge bin.

  9. Full-scale demonstration Low-NO sub x Cell trademark Burner retrofit

    SciTech Connect (OSTI)

    Not Available

    1992-05-11T23:59:59.000Z

    The Low-NO{sub x} Cell{trademark} Burner operates on the principle of staged combustion. The lower burner of each two-nozzle cell is modified to accommodate all the fuel input previously handled by two nozzles. Secondary air, less than theoretically required for complete combustion, is introduced to the lower burner. The remainder of secondary air is directed to the upper port'' of each cell to complete the combustion process. B W/EPRI have thoroughly tested the LNCB{trademark} at two pilot scales (6 million Btu per hour and 100 million Btu per hour), and tested a single full-scale burner in a utility boiler. Combustion tests at two scales have confirmed NO{sub x} reduction with the low-NO{sub x} cell on the order of 50% relative to the standard cell burner at optimum operating conditions. The technology is now ready for full unit, full-scale demonstration.

  10. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    SciTech Connect (OSTI)

    Wachs, G.W.

    1997-11-01T23:59:59.000Z

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78).

  11. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    SciTech Connect (OSTI)

    Kurt D. Hamman; Ray A. Berry

    2008-09-01T23:59:59.000Z

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  12. DEVELOPMENT OF A MULTI-LOOP FLOW AND HEAT TRANSFER FACILITY FOR ADVANCED NUCLEAR REACTOR THERMAL HYDRAULIC AND HYBRID ENERGY SYSTEM STUDIES

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-09-01T23:59:59.000Z

    A new high-temperature multi-fluid, multi-loop test facility for advanced nuclear applications is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Molten salts have been identified as excellent candidate heat transport fluids for primary or secondary coolant loops, supporting advanced high temperature and small modular reactors (SMRs). Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed. A preliminary design configuration will be presented, with the required characteristics of the various components. The loop will utilize advanced high-temperature compact printed-circuit heat exchangers (PCHEs) operating at prototypic intermediate heat exchanger (IHX) conditions. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4) flow loop operating at low pressure (0.2 MPa) at a temperature of ~450°C. Experiment design challenges include identification of suitable materials and components that will withstand the required loop operating conditions. Corrosion and high temperature creep behavior are major considerations. The facility will include a thermal energy storage capability designed to support scaled process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will also provide important data for code ve

  13. LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013

    SciTech Connect (OSTI)

    Hallbert, Bruce; Thomas, Ken

    2014-07-01T23:59:59.000Z

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  14. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    SciTech Connect (OSTI)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04T23:59:59.000Z

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  15. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    SciTech Connect (OSTI)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31T23:59:59.000Z

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

    SciTech Connect (OSTI)

    Not Available

    1980-04-18T23:59:59.000Z

    This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

  17. Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

    SciTech Connect (OSTI)

    Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

    2009-08-01T23:59:59.000Z

    The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation. As an initial step in accomplishing this effort, the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies was held March 20–21, 2009, in Columbus, Ohio, to enable industry stakeholders and researchers in identification of the nuclear industry’s needs in the areas of future I&C technologies and corresponding technology gaps and research capabilities. Approaches for collaboration to bridge or fill the technology gaps were presented and R&D activities and priorities recommended. This report documents the presentations and discussions of the workshop and is intended to serve as a basis for the plan under development to achieve the goals of the I&C research pathway.

  18. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01T23:59:59.000Z

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

  19. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08T23:59:59.000Z

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  20. Establishing criteria for the design of a combination parallel and cross-flaming covered burner 

    E-Print Network [OSTI]

    Stark, Christopher Charles

    2003-01-01T23:59:59.000Z

    with the data for the temperatures observed. The areas under the curve, above 100 degrees C and within an exposure time boundary were used to compute utilization factors. The utilization factors provided a relative comparison of burner efficiency...

  1. Sandia Energy - Consortium for Advanced Simulation of Light Water...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Consortium for Advanced Simulation of Light Water Reactors (CASL) Home Stationary Power Nuclear Fuel Cycle Advanced Nuclear Energy Consortium for Advanced Simulation of Light Water...

  2. Low No sub x /SO sub x burner retrofit for utility cyclone boilers

    SciTech Connect (OSTI)

    Moore, K.; Martin, L.; Smith, J.

    1991-05-01T23:59:59.000Z

    The Low NO{sub x}/SO{sub x} (LNS) Burner Retrofit for Utility Cyclone Boilers program consists of the retrofit and subsequent demonstration of the technology at Southern Illinois Power Cooperative's (SIPC's) 33-MW unit 1 cyclone boiler located near Marion, Illinois. The LNS Burner employs a simple innovative combustion process burning high-sulfur Illinois coal to provide substantial SO{sub 2} and NO{sub x} control within the burner. A complete series of boiler performance and characterization tests, called the baseline tests, was conducted in October 1990 on unit 1 of SIPC's Marion Station. The primary objective of the baseline test was to collect data from the existing plant that could provide a comparison of performance after the LNS Burner retrofit. These data could confirm the LNS Burner's SO{sub x} and NO{sub x} emissions control and any effect on boiler operation. Further, these tests would provide to the project experience with the operating characteristics of the host unit as well as engineering design information to minimize technical uncertainties in the application of the LNS Burner technology.

  3. Low NOx Burner Design and Analysis for Conceptual Design of Oxygen-Based PC Boiler

    SciTech Connect (OSTI)

    Andrew Seltzer

    2005-05-01T23:59:59.000Z

    The objective of the low NOx burner design and analysis task of the Conceptual Design of Oxygen-Based PC Boiler study is to optimize the burner design to ensure stable ignition, to provide safe operation, and to minimize pollutant formation. The burners were designed and analyzed using the Fluent computer program. Four burner designs were developed: (1) with no over-fire gas (OFG) and 65% flue gas recycle, (2) with 20% OFG and 65% flue gas recycle, (3) with no OFG and 56% flue gas recycle and (4) with 20% OFG and 56% flue gas recycle. A 3-D Fluent simulation was made of a single wall-fired burner and horizontal portion of the furnace from the wall to the center. Without primary gas swirl, coal burnout was relatively small, due to the low oxygen content of the primary gas stream. Consequently, the burners were modified to include primary gas swirl to bring the coal particles in contact with the secondary gas. An optimal primary gas swirl was chosen to achieve sufficient burnout.

  4. Study of the effects of ambient conditions upon the performance of fam powered, infrared, natural gas burners

    SciTech Connect (OSTI)

    Bai, Tiejun

    1996-10-01T23:59:59.000Z

    The objective of this investigation is to characterize the operation of a fan powered infrared burner (PIR burner) at various gas compositions and ambient conditions and develop design guidelines for appliances containing PIR burners for satisfactory performance. The fan powered infrared burner is a technology introduced more recently in the residential and commercial markets. It is a surface combustor that elevates the temperature of the burner head to a radiant condition. A variety of metallic and ceramic materials are used for the burner heads. It has been demonstrated that infrared burners produce low CO and NO{sub x} emissions in a controlled geometric space. This project consists of both experimental research and numerical analysis. To conduct the experiments, an experimental setup has been developed and installed in the Combustion Laboratory at Clerk Atlanta University (CAU). This setup consists of a commercial deep fat fryer that has been modified to allow in-situ radiation measurements on the surface of the infrared burner via a view port installed on the side wall of the oil vat. Proper instrumentation including fuel/air flow rate measurement, exhaust gas emission measurement, and radiation measurement has been developed. The project is progressing well. The scheduled tasks for this period of time were conducted smoothly. Specifically: 1. Baseline experimental study at CAU has been completed. The data are now under detailed analysis and will be reported in next quarterly report. 2. Theoretical formulation and analysis of the PIR burner performance model are continuing. Preliminary results have been obtained.

  5. 2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012–October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Noncompliance issues • Discussion of the facility’s environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  6. 2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    mike lewis

    2011-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Discussion of the facility’s environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  7. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-09-01T23:59:59.000Z

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  8. A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design

    SciTech Connect (OSTI)

    Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

    2010-11-01T23:59:59.000Z

    Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

  9. Rapid starting methanol reactor system

    DOE Patents [OSTI]

    Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

    1984-01-01T23:59:59.000Z

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

  10. Performance of a high efficiency advanced coal combustor

    SciTech Connect (OSTI)

    Toqan, M.A.; Paloposki, T.; Yu, T.; Teare, J.D.; Beer, J.M. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

    1989-12-01T23:59:59.000Z

    Under contract from DOE-PETC, Combustion Engineering, Inc. undertook the lead-role in a multi-task R D program aimed at development of a new burner system for coal-based fuels; the goal was that this burner system should be capable of being retrofitted in oil- or gas-fired industrial boilers, or usable in new units. In the first phase of this program a high efficiency advanced coal combustor was designed jointly by CE and MIT. Its burner is of the multiannular design with a fixed shrouded swirler in the center immediately surrounding the atomizer gun to provide the primary act,'' and three further annuli for the supply of the secondary air.'' The degree of rotation (swirl) in the secondary air is variable. The split of the combustion air into primary and secondary air flows serves the purpose of flame stabilization and combustion staging, the latter to reduce NO{sub x} formation.

  11. Achieving New Source Performance Standards (NSPS) Emission Standards Through Integration of Low-NOx Burners with an Optimization Plan for Boiler Combustion

    SciTech Connect (OSTI)

    Wayne Penrod

    2006-12-31T23:59:59.000Z

    The objective of this project was to demonstrate the use of an Integrated Combustion Optimization System to achieve NO{sub X} emission levels in the range of 0.15 to 0.22 lb/MMBtu while simultaneously enabling increased power output. The project plan consisted of the integration of low-NO{sub X} burners and advanced overfire air technology with various process measurement and control devices on the Holcomb Station Unit 1 boiler. The plan included the use of sophisticated neural networks or other artificial intelligence technologies and complex software to optimize several operating parameters, including NO{sub X} emissions, boiler efficiency, and CO emissions. The program was set up in three phases. In Phase I, the boiler was equipped with sensors that can be used to monitor furnace conditions and coal flow to permit improvements in boiler operation. In Phase II, the boiler was equipped with burner modifications designed to reduce NO{sub X} emissions and automated coal flow dampers to permit on-line fuel balancing. In Phase III, the boiler was to be equipped with an overfire air system to permit deep reductions in NO{sub X} emissions. Integration of the overfire air system with the improvements made in Phases I and II would permit optimization of boiler performance, output, and emissions. This report summarizes the overall results from Phases I and II of the project. A significant amount of data was collected from the combustion sensors, coal flow monitoring equipment, and other existing boiler instrumentation to monitor performance of the burner modifications and the coal flow balancing equipment.

  12. Study of the effects of ambient conditions upon the performance of fan powered, infrared, natural gas burners. Quarterly technical progress report, October 1, 1996--December 31, 1996

    SciTech Connect (OSTI)

    Bai, T.

    1997-01-01T23:59:59.000Z

    This quarterly technical progress report describes work performed under DOE Grant No. DE-FG22-94MT94011 during the period September 1, 1996 to December 31, 1996 which covers the nineth quarter of the project. The objective of this investigation is to characterize the operation of a fan powered infrared burner (IR burner) at various gas compositions and ambient conditions and develop design guidelines for appliances containing PIR burners for satisfactory performance. The fan powered infrared burner is a technology introduced more recently in the residential and commercial markets. It is a surface combustor that elevates the temperature of the burner head to a radiant condition. A variety of metallic and ceramic materials are used for the burner heads. It has been demonstrated that infrared burners produce low CO and NO{sub x} emissions in a controlled geometric space. As the environmental regulations become more stringent, infrared burners are receiving increasing interests.

  13. Thermionic-cogeneration-burner assessment study. Second quarterly technical progress report, January-March 1983

    SciTech Connect (OSTI)

    Not Available

    1983-01-01T23:59:59.000Z

    The performance analysis work continued with the completion of the programming of the mathematical model and with the start of a series of parametric analyses. Initial studies predict that approximately 25 to 30% of the heat contained in the flue gas can be passed through the thermionic converters (TEC) and then be converted at 12 to 15% efficiency into electrical power. This results in up to 17 kWe per 1 million Btu/h burner firing rate. This is a 4 to 10 percent energy saving over power produced at the utility. The thermal burner design and construction have been completed, as well as initial testing on the furnace and preheat systems. The following industries are still considered viable options for use of the thermionic cogeneration burner: chlor-alkali, alumina-aluminum, copper refining, steel and gray iron, industries using resistance heating, electrolytic industries and electrochemical industries. Information gathered on these industries is presented.

  14. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

    Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

    2005-01-01T23:59:59.000Z

    The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

  15. Gas-phase and catalytic combustion in heat-recirculating burners Jeongmin Ahn, Craig Eastwood, Lars Sitzki* and Paul D. Ronney

    E-Print Network [OSTI]

    1 Gas-phase and catalytic combustion in heat-recirculating burners Jeongmin Ahn, Craig Eastwood title: Extinction limits in excess enthalpy burners To be published in Proceedings of the Combustion-phase and catalytic combustion in heat-recirculating burners Jeongmin Ahn, Craig Eastwood, Lars Sitzki* and Paul D

  16. Development and validation of a combustion model for a fuel cell off-gas burner

    E-Print Network [OSTI]

    Collins, William Tristan

    2008-10-14T23:59:59.000Z

    Development and Validation of a Combustion Model for a Fuel Cell Off-Gas Burner W. Tristan Collins Magdalene College University of Cambridge A dissertation submitted to the University of Cambridge for the degree of Doctor of Philosophy June 2008... Development and Validation of a Combustion Model for a Fuel Cell Off-Gas Burner W. Tristan Collins A low-emissions power generator comprising a solid oxide fuel cell coupled to a gas turbine has been developed by Rolls-Royce Fuel Cell Systems. As part...

  17. Sandia National Laboratories: Small Modular Reactors

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    advanced computing capabilities that serve as a virtual version of existing, operating nuclear reactors-to enable nuclear energy to continue to provide dependable, afford-able...

  18. TURBULENT COMBUSTION MODELING OF COAL:BIOMASS BLENDS IN A SWIRL BURNER I -PRELIMINARY RESULTS

    E-Print Network [OSTI]

    Daripa, Prabir

    coal or by ex- haust clean up technology. For the power plants, the simplest solution is the preventive- ity well into the 21st century. This dependency on coal calls for better technologies to reduceTURBULENT COMBUSTION MODELING OF COAL:BIOMASS BLENDS IN A SWIRL BURNER I - PRELIMINARY RESULTS

  19. Low NO sub x /SO sub x Burner retrofit for utility cyclone boilers

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    Cyclone furnaces operate with high excess air and at high temperature. The heat release during combustion is very high and as a result the boiler volume is much smaller than would be found in a conventional pc-fired system. The Marion Unit 1 boiler, at the level of the cyclone entry, has a small cross-section; about 5-feet in depth and about 20-feet in width. A boiler schematic showing the LNS Burner and relative location of the superheater region and overfire air ports is shown in Figure 1. The LNS Burner's combustion process is fundamentally different from that of the cyclone, and the combustion products are also different. The LNS Burner products enter the boiler as hot, fuel-rich gases. Additional overfire air must be added to complete this combustion step with care taken to avoid the formation of thermal NO{sub x}. If done correctly, S0{sub 2} is controlled and significant NO{sub x} reductions are achieved. Because of the small boiler volume, flow modelling was found to be necessary to insure that adequate mixing of LNS Burner combustion products with air can be accomplished to achieve NO{sub x} emissions goals. Design requirements for the air injection system for the Marion boiler were developed using FLUENT, a commercially available computational fluid dynamics (CFD) code. A series of runs were made to obtain a design for final air injection that met the process design goals as closely as possible.

  20. Establishing criteria for the design of a combination parallel and cross-flaming covered burner

    E-Print Network [OSTI]

    Stark, Christopher Charles

    2003-01-01T23:59:59.000Z

    it with the two open flame practices. This evaluation was performed by moving the burners over an area that would monitor the temperatures at specified heights and locations. Temperatures were measured using thermocouples placed at heights 7-mm, 150-mm, and 300...